SEC Petition Evaluation Report 

Petition SEC-00126 

Report Rev #: 1 Report Submittal Date: September 14, 2009

Subject Expert(s): Louise Buker, Roger Halsey, Karin Jessen, Dan Mantooth, Eugene W. Potter 

Site Expert(s): N/A 

Petition Administrative Summary 

Petition Under Evaluation 

Petition # Petition 

Type 

Petition 

Qualification Date 

DOE/AWE Facility Name 

SEC-00126 83.13 December 22, 2008 Piqua Organic Moderated Reactor 

Petitioner Class Definition 

All employees associated with reactor activities who worked within and around the reactor dome at the Piqua Organic 

Moderated Reactor during the covered period from January 1, 1963 through December 31, 1966. 

Class Evaluated by NIOSH 

All employees associated with reactor activities who worked within and around the reactor dome at the Piqua Organic 

Moderated Reactor during the covered period from January 1, 1963 through February 28, 1969. 

NIOSH-Proposed Class to be Added to the SEC 

All employees of the Department of Energy, its predecessor agencies, and its contractors and subcontractors who worked 

at the Piqua Organic Moderated Reactor site during the covered period from May 2, 1966 through February 28, 1969, for 

a number of work days aggregating at least 250 work days, occurring either solely under this employment or in 

combination with work days within the parameters established for one or more other classes of employees in the Special 

Exposure Cohort.

Related Petition Summary Information 

SEC Petition Tracking #(s) Petition Type DOE/AWE Facility Name Petition Status 

N/A N/A N/A N/A 

Related Evaluation Report Information 

Report Title DOE/AWE Facility Name 

N/A N/A 

ORAU Lead Technical Evaluator: Karin Jessen ORAU Review Completed By: Daniel Stempfley 

Peer Review Completed By: [Signature on file] 9/24/2009 

Charles D. Nelson Date 

SEC Petition Evaluation Reviewed By: [Signature on file] 9/24/2009 

J. W. Neton Date 

SEC Evaluation Approved By: [Signature on file] 9/24/2009 

Larry Elliott Date 
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Evaluation Report Summary: SEC-00126, Piqua Organic Moderated Reactor 

This evaluation report by the National Institute for Occupational Safety and Health (NIOSH) 

addresses a class of employees proposed for addition to the Special Exposure Cohort (SEC) per the 

Energy Employees Occupational Illness Compensation Program Act of 2000, as amended, 42 U.S.C. 

§ 7384 et seq. (EEOICPA) and 42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as 

Members of the Special Exposure Cohort under the Energy Employees Occupational Illness 

Compensation Program Act of 2000. 

Petitioner-Requested Class Definition

Although the petitioner initially requested a wider date range for the petitioner-requested class, after 

consulting with the petitioner, the class was limited to the covered period, 1963 through 1966, as 

listed by the Department of Energy (DOE) Office of Health, Safety and Security at the time of 

qualification. 

Petition SEC-00126, qualified on December 22, 2008, requested that NIOSH consider the following 

class: All employees associated with reactor activities who worked within and around the reactor 

dome at the Piqua Organic Moderated Reactor during the covered period from January 1, 1963 

through December 31, 1966. 

Class Evaluated by NIOSH

Based on its preliminary research and the DOE/Department of Labor’s (DOL) decision to expand the 

covered period for the POMR site, NIOSH expanded the petitioner-requested class to include the postoperational period, which includes the recovery period and decontamination and decommissioning 

activities that were completed February 28, 1969. NIOSH evaluated the following class: All 

employees associated with reactor activities who worked within and around the reactor dome at the 

Piqua Organic Moderated Reactor during the covered period from January 1, 1963 through February 

28, 1969. 

NIOSH-Proposed Class to be Added to the SEC

Based on its full research of the class under evaluation, NIOSH has defined a single class of 

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class includes: All employees of the Department of Energy, its predecessor agencies, and its 

contractors and subcontractors who worked at the Piqua Organic Moderated Reactor site during the 

covered period from May 2, 1966 through February 28, 1969, for a number of work days aggregating 

at least 250 work days, occurring either solely under this employment or in combination with work 

days within the parameters established for one or more other classes of employees in the Special 

Exposure Cohort. NIOSH is proposing only the post-operational period based on (1) the more 

extensive personnel exposure potential during the post-operational period (based on the recovery and 

decontamination and decommissioning activities performed), and (2) the lack of personnel internal 

monitoring data for the post-operational period. 
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Feasibility of Dose Reconstruction

NIOSH finds it is not feasible to estimate radiation exposures with sufficient accuracy for all workers 

at the POMR site during the post-operational period from May 2, 1966 through February 28, 1969. 

With the exception of this class, per EEOICPA and 42 C.F.R. § 83.13(c)(1), NIOSH has established 

that it has access to sufficient information to: (1) estimate the maximum radiation dose, for every type 

of cancer for which radiation doses are reconstructed, that could have been incurred in plausible 

circumstances; or (2) estimate radiation doses more precisely than an estimate of maximum dose. 

Information available to NIOSH is sufficient to document or estimate the maximum internal and 

external potential exposure to members of the evaluated class under plausible circumstances during 

the specified operational period (January 1, 1963 through May 1, 1966). 

Health Endangerment Determination

Per EEOICPA and 42 C.F.R. § 83.13(c)(3), a health endangerment determination is required because 

NIOSH has determined that it does not have sufficient information to estimate dose for the members 

of the proposed class during the post-operational period from May 2, 1966 through February 28, 1969. 

NIOSH did not identify any evidence supplied by the petitioners or from other resources that would 

establish that the proposed class was exposed to radiation during a discrete incident likely to have 

involved exceptionally high-level exposures. However, evidence indicates that some workers in the 

proposed class may have accumulated substantial chronic exposures through episodic intakes of 

radionuclides, combined with external exposures to gamma, beta, and neutron radiation. 

Consequently, NIOSH has determined that health was endangered for those workers covered by this 

evaluation who were employed for at least 250 aggregated work days either solely under their 

employment or in combination with work days within the parameters established for other SEC 

classes (excluding aggregate work day requirements). 

For the period January 1, 1963 through May 1, 1966, a health endangerment determination is not 

required because NIOSH has determined that it has sufficient information to estimate dose for the 

members of the evaluated class. 
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Table of Contents 

Evaluation Report Summary: SEC-00126, Piqua Organic Moderated Reactor...................................................................... 3 

1.0 Purpose and Scope ....................................................................................................................................................... 9 

2.0 Introduction.................................................................................................................................................................. 9 

3.0 SEC-00126, Piqua Organic Moderated Reactor Class Definitions ............................................................................ 10

3.1 Petitioner-Requested Class Definition and Basis ............................................................................................ 10

3.2 Class Evaluated by NIOSH ............................................................................................................................. 11

3.3 NIOSH-Proposed Class to be Added to the SEC ............................................................................................ 12 

4.0 Data Sources Reviewed by NIOSH to Evaluate the Class ......................................................................................... 12

4.1 Site Profile Technical Basis Documents (TBDs) ............................................................................................ 12

4.2 ORAU Technical Information Bulletin (OTIB) .............................................................................................. 13

4.3 Facility Employees and Experts ...................................................................................................................... 13

4.4 Previous Dose Reconstructions ....................................................................................................................... 14

4.5 NIOSH Site Research Database ...................................................................................................................... 14

4.6 Documentation and/or Affidavits Provided by Petitioners.............................................................................. 14 

5.0 Radiological Operations Relevant to the Class Evaluated by NIOSH ....................................................................... 15

5.1 POMR Plant and Process Descriptions............................................................................................................ 15

5.1.1 POMR Plant and Process Descriptions for the Operational Period................................................... 15

5.1.2 POMR Plant and Process Descriptions for the Post-Operational Period........................................... 23

5.1.2.1 Recovery Period .............................................................................................................. 25

5.1.2.2 Decontamination and Decommissioning......................................................................... 27

5.2 Internal Radiological Exposure Sources from Operational and Post-Operational POMR Activities .............. 29

5.2.1 Activated Impurities in the Coolant .................................................................................................. 31

5.2.2 Activated Corrosion Products ........................................................................................................... 31

5.2.3 Neutron Recoil Reaction Products .................................................................................................... 32

5.2.4 Fission Products ................................................................................................................................ 32

5.3 External Radiological Exposure Sources from Operational and Post-Operational POMR Activities............. 33

5.3.1 Photon ............................................................................................................................................... 34

5.3.2 Beta ................................................................................................................................................... 34

5.3.3 Neutron.............................................................................................................................................. 35 

6.0 Summary of Available Monitoring Data for the Class Evaluated by NIOSH............................................................ 35

6.1 Available POMR Internal Monitoring Data .................................................................................................... 35

6.1.1 Available POMR Internal Monitoring Data for the Operational Period ........................................... 36

6.1.2 Available POMR Internal Monitoring Data for the Post-Operational Period ................................... 36

6.2 Available POMR External Monitoring Data ................................................................................................... 36

6.2.1 Available POMR External Monitoring Data for the Operational Period .......................................... 37

6.2.2 Available POMR External Monitoring Data for the Post-Operational Period .................................. 41 

7.0 Feasibility of Dose Reconstruction for the Class Evaluated by NIOSH .................................................................... 43

7.1 Pedigree of POMR Data.................................................................................................................................. 44

7.1.1 Internal Monitoring Data Pedigree Review....................................................................................... 44

7.1.1.1 Internal Monitoring Data Pedigree Review for the Operational Period .......................... 44

7.1.1.2 Internal Monitoring Data Pedigree Review for the Post-Operational Period .................. 45

7.1.2 External Monitoring Data Pedigree Review...................................................................................... 45

7.1.2.1 External Monitoring Data Pedigree Review for the Operational Period ......................... 45

7.1.2.2 External Monitoring Data Pedigree Review for the Post-Operational Period................. 46

7.2 Evaluation of Bounding Internal Radiation Doses at POMR .......................................................................... 46

7.2.1 Evaluation of Bounding Internal Doses during the Operational Period ............................................ 47

7.2.1.1 Evaluation of Bounding Process-Related Internal Doses during the Operational 
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 Period .............................................................................................................................. 47

7.2.1.2 Evaluation of Bounding Ambient Environmental Internal Doses during the Operational 

Period .............................................................................................................................. 48

7.2.1.3 Methods for Bounding Process-Related Internal Doses during the Operational 

 Period .............................................................................................................................. 48

7.2.1.4 Methods for Bounding Ambient Internal Doses during the Operational Period ............. 50

7.2.1.5 Internal Dose Reconstruction Feasibility Conclusion for the Operational Period........... 50

7.2.2 Evaluation of Bounding Internal Doses during the Post-Operational Period.................................... 51

7.2.2.1 Evaluation of Bounding Process-Related Internal Doses during the Post-Operational 

Period .............................................................................................................................. 51

7.2.2.2 Evaluation of Bounding Ambient Environmental Internal Doses during the PostOperational Period........................................................................................................... 52

7.2.2.3 Methods for Bounding Process-Related Internal Doses during the Post-Operational 

Period .............................................................................................................................. 52

7.2.2.4 Methods for Bounding Ambient Internal Doses during the Post-Operational Period ..... 53

7.2.2.5 Internal Dose Reconstruction Feasibility Conclusion for the Post-Operational 

 Period .............................................................................................................................. 53

7.3 Evaluation of Bounding External Radiation Doses at POMR ......................................................................... 53

7.3.1 Evaluation of Bounding External Doses during the Operational Period........................................... 53

7.3.1.1 Personnel Dosimetry Data for the Operational Period .................................................... 53

7.3.1.2 Area Monitoring Data for the Operational Period........................................................... 55

7.3.1.3 POMR Occupational X-Ray Examinations..................................................................... 56

7.3.1.4 Evaluation for Bounding Ambient External Doses during the Operational Period......... 56

7.3.1.5 Methods for Bounding External Doses during the Operational Period ........................... 56

7.3.1.6 External Dose Reconstruction Feasibility Conclusion for the Operational Period.......... 58

7.3.2 Evaluation of Bounding External Doses during the Post-Operational Period................................... 58

7.3.2.1 Personnel Dosimetry Data for the Post-Operational Period ............................................ 59

7.3.2.2 Area Monitoring Data for the Post-Operational Period................................................... 60

7.3.2.3 POMR Occupational X-Ray Examinations..................................................................... 61

7.3.2.4 Evaluation for Bounding Ambient External Doses during the Post-Operational 

 Period .............................................................................................................................. 61

7.3.2.5 Methods for Bounding External Doses during the Post-Operational Period................... 61

7.3.2.6 External Dose Reconstruction Feasibility Conclusion for the Post-Operational 

 Period .............................................................................................................................. 62

7.4 Evaluation of Petition Basis for SEC-00126 ................................................................................................... 62

7.5 Summary of Feasibility Findings for Petition SEC-00126.............................................................................. 62 

8.0 Evaluation of Health Endangerment for Petition SEC-00126.................................................................................... 63 

9.0 Class Conclusion for Petition SEC-00126 ................................................................................................................. 64 

10.0 References.................................................................................................................................................................. 65 

Attachment One: Information from Interviews with Former Piqua Personnel ..................................................................... 71

Attachment Two: Data Capture Synopsis............................................................................................................................. 78
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Tables 

Table 4-1: No. of POMR Claims Submitted Under the Dose Reconstruction Rule.............................................................. 14 

Table 5-1: Summary of POMR Operational Period History................................................................................................. 16

Table 5-2: Summary of POMR Post-Operational Period History......................................................................................... 24

Table 5-3: Radionuclides of Concern for POMR Operational and Post-Operational Internal Exposures ............................ 33 

Table 6-1: Summary of Annual Whole-Body Penetrating Exposures for the POMR Operational Period............................ 37

Table 6-2: Semiannual and Monthly Report Results for the POMR Operational Period ..................................................... 38

Table 6-3: Summary of Annual Whole-Body Penetrating Exposures for the POMR Post-Operational Period ................... 41

Table 6-4: Semiannual and Monthly Report Results for the POMR Post-Operational Period ............................................. 42 

Table 7-1: Radionuclides of Concern by Category............................................................................................................... 49

Table 7-2: Summary of Feasibility Findings for SEC-00126 ............................................................................................... 63

Figures 

Figure 5-1: Piqua Core I Fuel Element ................................................................................................................................. 19 
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SEC Petition Evaluation Report for SEC-00126

1.0 Purpose and Scope 

This report evaluates the feasibility of reconstructing doses for all employees associated with reactor 

activities who worked within and around the reactor dome at the Piqua Organic Moderated Reactor 

during the covered period from January 1, 1963 through February 28, 1969. It provides information 

and analyses germane to considering a petition for adding a class of employees to the congressionallycreated SEC. 

This report does not make any determinations concerning the feasibility of dose reconstruction that 

necessarily apply to any individual energy employee who might require a dose reconstruction from 

NIOSH. This report also does not contain the final determination as to whether the proposed class 

will be added to the SEC (see Section 2.0). 

This evaluation was conducted in accordance with the requirements of EEOICPA, 42 C.F.R. pt. 83, 

and the guidance contained in the Office of Compensation Analysis and Support’s (OCAS) Internal 

Procedures for the Evaluation of Special Exposure Cohort Petitions, OCAS-PR-004. 

2.0 Introduction 

Both EEOICPA and 42 C.F.R. pt. 83 require NIOSH to evaluate qualified petitions requesting that the 

Department of Health and Human Services (HHS) add a class of employees to the SEC. The 

evaluation is intended to provide a fair, science-based determination of whether it is feasible to 

estimate with sufficient accuracy the radiation doses of the class of employees through NIOSH dose 

reconstructions.1 

42 C.F.R. § 83.13(c)(1) states: Radiation doses can be estimated with sufficient accuracy if NIOSH 

has established that it has access to sufficient information to estimate the maximum radiation dose, 

for every type of cancer for which radiation doses are reconstructed, that could have been incurred in 

plausible circumstances by any member of the class, or if NIOSH has established that it has access to 

sufficient information to estimate the radiation doses of members of the class more precisely than an 

estimate of the maximum radiation dose.

 

Under 42 C.F.R. § 83.13(c)(3), if it is not feasible to estimate with sufficient accuracy radiation doses 

for members of the class, then NIOSH must determine that there is a reasonable likelihood that such 

radiation doses may have endangered the health of members of the class The regulation requires 

NIOSH to assume that any duration of unprotected exposure may have endangered the health of 

members of a class when it has been established that the class may have been exposed to radiation 

during a discrete incident likely to have involved levels of exposure similarly high to those occurring 

during nuclear criticality incidents. If the occurrence of such an exceptionally high-level exposure has 

not been established, then NIOSH is required to specify that health was endangered for those workers 

1

 NIOSH dose reconstructions under EEOICPA are performed using the methods promulgated under 42 C.F.R. pt. 82 and 

the detailed implementation guidelines available at http://www.cdc.gov/niosh/ocas. 
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who were employed for at least 250 aggregated work days within the parameters established for the 

class or in combination with work days within the parameters established for other SEC classes 

(excluding aggregate work day requirements). 

NIOSH is required to document its evaluation in a report, and to do so, relies upon both its own dose 

reconstruction expertise as well as technical support from its contractor, Oak Ridge Associated 

Universities (ORAU). Once completed, NIOSH provides the report to both the petitioner(s) and to the 

Advisory Board on Radiation and Worker Health (Board). The Board will consider the NIOSH 

evaluation report, together with the petition, petitioner(s) comments, and other information the Board 

considers appropriate, in order to make recommendations to the Secretary of HHS on whether or not 

to add one or more classes of employees to the SEC. Once NIOSH has received and considered the 

advice of the Board, the Director of NIOSH will propose a decision on behalf of HHS. The Secretary 

of HHS will make the final decision, taking into account the NIOSH evaluation, the advice of the 

Board, and the proposed decision issued by NIOSH. As part of this decision process, petitioners may 

seek a review of certain types of final decisions issued by the Secretary of HHS.2

 

3.0 SEC-00126, Piqua Organic Moderated Reactor Class Definitions 

The following subsections address the evolution of the class definition for SEC-00126, Piqua Organic 

Moderated Reactor (POMR). When a petition is submitted, the requested class definition is reviewed 

as submitted. Based on its review of the available site information and data, NIOSH will make a 

determination whether to qualify for full evaluation all, some, or no part of the petitioner-proposed 

class. If some portion of the petitioner-proposed class is qualified, NIOSH will specify that class 

along with a justification for any modification of petitioner’s class. After a full evaluation of the 

qualified class, NIOSH will determine whether to propose a class for addition to the SEC and will 

specify that proposed class definition. 

3.1 Petitioner-Requested Class Definition and Basis 

The petitioner initially requested a wider date range for the petitioner-requested class; specifically, 

Form B included two periods: December 21, 1959 through January 28, 1967 and October 19, 1967 

through January 30, 1987. After consulting the petitioner, the class was limited to the covered period, 

1963 through 1966, as listed by the Department of Energy (DOE) Office of Health, Safety and 

Security at the time of qualification. 

Petition SEC-00126, qualified on December 22, 2008, requested that NIOSH consider the following 

class for addition to the SEC: All employees associated with reactor activities who worked within and 

around the reactor dome at the Piqua Organic Moderated Reactor during the covered period from 

January 1, 1963 through December 31, 1966.

The petitioner provided information and affidavit statements in support of the petitioner’s belief that 

accurate dose reconstruction over time is impossible for the POMR workers in question. NIOSH 

deemed the following information and affidavit statements sufficient to qualify SEC-00126 for 

evaluation: 

2

 See 42 C.F.R. pt. 83 for a full description of the procedures summarized here. Additional internal procedures are 

available at http://www.cdc.gov/niosh/ocas. 
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In support of the petition, the SEC-00126 petitioner claims that no records were kept on 

activities related to the Piqua site to include the dismantling of the facility and that the 

petitioner’s father was not trained in the hazards of the POMR. In addition, no monitoring 

devices were ever offered.

During the construction and operation of the reactor, [former POMR employee] was 

employed by the city of Piqua as a material handler and subsequently a linesman for the 

power plant and helped out at the reactor site at various times even though he was primarily 

assigned to the original power plant adjacent to one another. He was a laborer primarily 

assigned any duty his supervisor would assign. This all done without any type of monitoring, 

training, or protective devices for handling nuclear material [sic] this was not explained but 

one has to surmise it was because of [former POMR employee]’s primary work was at the 

other building [sic]. It was during the second part of his career, decommissioning and 

dismantling of the Reactor, that [former POMR employee] had the most potential exposure to 

the harmful elements used at and remain at the Moderated Reactor located in Piqua Ohio. 

... At no time was [name of the former Energy employee] trained in the hazards [at the POMR 

site] nor was any monitoring devices offered.” “No records were kept on activities related to 

the Piqua site to include the dismantling of the facility and use of [former POMR employee 

and survivor’s father] in such exposure. 

Based on its POMR research and data capture efforts, NIOSH determined that it has access to 

monthly, semiannual, and annual summary reports, reactor design, shielding material, and radiation 

source information for the time period under evaluation, but NIOSH has also determined that internal 

and external records are not complete for all time periods or for all radionuclides. The information 

and statements provided by the petitioner qualified the petition for further consideration by NIOSH, 

the Board, and HHS. 

3.2 Class Evaluated by NIOSH 

The covered time period for the POMR site was extended by DOE/DOL so that the covered period 

included the post-operational period in which decontamination and decommissioning activities were 

performed. The DOE EEOICPA Database now lists the covered time period for the POMR site from 

1963 to 1969. 

Based on its preliminary research and the DOE/Department of Labor’s decision to expand the covered 

period for the POMR site, NIOSH expanded the petitioner-proposed class to include the postoperational period, which includes the recovery period and the decontamination and decommissioning 

activities that were completed February 28, 1969. Therefore, NIOSH defined the following class for 

further evaluation: All employees associated with reactor activities who worked within and around the 

reactor dome at the Piqua Organic Moderated Reactor during the covered period from January 1, 1963 

through February 28, 1969. 
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3.3 NIOSH-Proposed Class to be Added to the SEC 

Based on its research of the class under evaluation, NIOSH has defined a single class of employees for 

which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSH-proposed class 

to be added to the SEC includes all employees of the Department of Energy, its predecessor agencies, 

and its contractors and subcontractors who worked at the Piqua Organic Moderated Reactor site 

during the covered period from May 2, 1966 through February 28, 1969, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment or in combination 

with work days within the parameters established for one or more other classes of employees in the 

Special Exposure Cohort. 

4.0 Data Sources Reviewed by NIOSH to Evaluate the Class 

In addition to searching City of Piqua records, NIOSH completed an extensive database and Internet 

search for information regarding the POMR facility. The database search included the DOE 

(Department of Energy) Legacy Management Considered Sites database, the DOE Office of Scientific 

and Technical Information (OSTI) database, the Energy Citations database, the Atomic Energy 

Technical Report database, and the Hanford Declassified Document Retrieval System. In addition to 

general Internet searches, the NIOSH Internet search included OSTI OpenNet Advanced searches, 

OSTI Information Bridge Fielded searches, Nuclear Regulatory Commission (NRC) Agency-wide 

Documents Access and Management (ADAMS) Web searches, the DOE Office of Human Radiation 

Experiments website, and the DOE-National Nuclear Security Administration-Nevada Site Officesearch. Attachment Two contains a summary of POMR documents. The summary specifically 

identifies data capture details and general descriptions of the documents retrieved. 

In addition to the database and Internet searches listed above, NIOSH identified and reviewed 

numerous data sources to determine information relevant to determining the feasibility of dose 

reconstruction for the class of employees under evaluation. This included determining the availability 

of information on personal monitoring, area monitoring, industrial processes, and radiation source 

materials. The following subsections summarize the data sources identified and reviewed by NIOSH. 

4.1 Site Profile Technical Basis Documents (TBDs) 

A Site Profile provides specific information concerning the documentation of historical practices at 

the specified site. Dose reconstructors can use the Site Profile to evaluate internal and external 

dosimetry data for monitored and unmonitored workers, and to supplement, or substitute for, 

individual monitoring data. A Site Profile consists of an Introduction and five Technical Basis 

Documents (TBDs) that provide process history information, information on personal and area 

monitoring, radiation source descriptions, and references to primary documents relevant to the 

radiological operations at the site. The Site Profile for a small site may consist of a single document. 

NIOSH has not prepared a Site Profile or TBD specific to the POMR site. 
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4.2 ORAU Technical Information Bulletin (OTIB) 

An ORAU Technical Information Bulletin (OTIB) is a general working document that provides 

guidance for preparing dose reconstructions at particular sites or categories of sites. NIOSH reviewed 

the following OTIB as part of its evaluation: 

 OTIB: Dose Reconstruction from Occupationally Related Diagnostic X-Ray Procedures, ORAUTOTIB-0006; December 21, 2005; SRDB Ref ID: 20220

4.3 Facility Employees and Experts 

To obtain additional information, NIOSH contacted nine former POMR facility personnel. NIOSH 

performed eleven interviews with nine former Piqua employees (two former employees were 

interviewed on two separate occasions), all whom were considered to be knowledgeable about the 

POMR facility. All interviews were conducted by phone. The purpose of the interviews was to gain 

additional first-hand information from people who worked at the POMR facility. A summary of the 

information obtained from the interviews can be found in Attachment One of this report. 

 Personal Communication, 2009a, Personal Communication with Former Health Physics 

Technician; Telephone Interview by ORAU Team; January 20, 2009; SRDB Ref ID: 61683

 Personal Communication, 2009b, Personal Communication with DOE Legacy Management 

Employee; Telephone Interview by ORAU Team; January 29, 2009; SRDB Ref ID: 61681 

 Personal Communication, 2009c, Personal Communication with Former Shift Supervisor; 

Telephone Interview by ORAU Team; February 19, 2009; SRDB Ref ID: 61677

 Personal Communication, 2009d, Personal Communication with Former Construction Engineer, 

Instrumentation Engineer, and Electrical Engineer; Telephone Interview by ORAU Team; 

February 19, 2009; SRDB Ref ID: 61684 

 Personal Communication, 2009e, Personal Communication with Former Reactor Operator and 

Maintenance Foreman; Telephone Interview by ORAU Team; February 6, 2009; SRDB Ref ID: 

61679

 Personal Communication, 2009f, Personal Communication with Former Chief Health Physicist; 

Telephone Interview by ORAU Team; February 23, 2009; SRDB Ref ID: 61680

 Personal Communication, 2009g, Personal Communication with Former Health Physics 

Technician, second interview; Telephone Interview by ORAU Team; March 18, 2009; SRDB Ref 

ID: 62597

 Personal Communication, 2009h, Personal Communication with Former POMR Health Physicist; 

documented telephone communication; March 23, 2009; SRDB Ref ID: 62596

 Personal Communication, 2009i, Personal Communication with Former Shift Supervisor; 

Telephone Interview by ORAU Team; June 23, 2009; SRDB Ref ID: 71375 
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 Personal Communication, 2009j, Personal Communication with Former Day Shift Supervisor, 

Operations Engineer, and Operations and Maintenance Supervisor; Telephone Interview by 

ORAU Team; June 29, 2009; SRDB Ref ID: 71376 

 Personal Communication, 2009k, Personal Communication with Former Health Physicist; 

Telephone Interview by ORAU Team; June 25, 2009; SRDB Ref ID: 71379 

4.4 Previous Dose Reconstructions 

NIOSH reviewed its NIOSH OCAS Claims Tracking System (NOCTS) to locate EEOICPA-related 

dose reconstructions that might provide information relevant to the petition evaluation. Table 4-1 

summarizes the results of this review. (NOCTS data available as of September 9, 2009) 

Table 4-1: No. of POMR Claims Submitted Under the Dose Reconstruction Rule 

Description Totals 

Total number of claims submitted for dose reconstruction 5 

Total number of claims submitted for energy employees who meet the definition criteria for the class 

under evaluation (January 1, 1963 through February 28, 1969) 5 

Number of dose reconstructions completed for energy employees who meet the definition criteria for 

the class under evaluation (i.e., the number of such claims completed by NIOSH and submitted to the 

Department of Labor for final approval). 3 

Number of claims for which internal dosimetry records were obtained for the identified years in the 

evaluated class definition 0 

Number of claims for which external dosimetry records were obtained for the identified years in the 

evaluated class definition 1 

NIOSH reviewed each claim to determine whether internal and/or external personal monitoring 

records could be obtained for the employee. Five claims have been submitted for the POMR facility; 

three of the claims have been completed. NIOSH has located no claims with internal monitoring and 

only one claim with external monitoring for individuals that were employed during the period 

evaluated in this report. 

4.5 NIOSH Site Research Database 

NIOSH also examined its Site Research Database (SRDB) to locate documents supporting the 

assessment of the evaluated class. Two hundred twenty-eight documents in this database were 

identified as pertaining to the POMR facility. These documents were evaluated for their relevance to 

this petition. The documents include personnel data, historical background information on the POMR 

facility, as well as monthly, semiannual, and annual reports. 

4.6 Documentation and/or Affidavits Provided by Petitioners 

In qualifying and evaluating the petition, NIOSH reviewed the following document submitted by the 

petitioner: 

 Petition Form B; August 21, 2008; OSA Ref ID: 106811 
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5.0 Radiological Operations Relevant to the Class Evaluated by 

NIOSH 

The following subsections summarize both radiological operations at the POMR facility from January 

1, 1963 through February 28, 1969 and the information available to NIOSH to characterize particular 

processes and radioactive source materials. From available sources NIOSH has gathered monthly, 

semiannual, and annual summary reports, as well as reactor design information, shielding information, 

and radiation source material information. The information included within this evaluation report is 

intended only to be a summary of the available information. 

5.1 POMR Plant and Process Descriptions 

ATTRIBUTION: Section 5.1and its related subsections were completed by Karin Jessen, Oak Ridge 

Associated Universities (ORAU). These conclusions were peer-reviewed by the individuals listed on 

the cover page. The rationales for all conclusions in this document are explained in the associated 

text.

From 1963 to 1966, the Piqua Nuclear Power Facility was operated as a demonstration project by the 

City of Piqua. The facility contained a 45.5-megawatt (thermal) organically cooled and moderated 

reactor. In 1966, the Atomic Energy Commission (AEC) discontinued facility operations and 

terminated its contract with the city. The AEC dismantled and decommissioned the reactor between 

1967 and 1969. The reactor fuel coolant and most of the radioactive materials were removed from the 

site. As referenced throughout this report, the POMR operational period is from January 1, 1963 

through May 1, 1966, and the post-operational period is from May 2, 1966 through February 28, 1969. 

5.1.1 POMR Plant and Process Descriptions for the Operational Period 

The POMR site, also known as the Piqua Nuclear Power facility, was a nuclear power plant designed 

by Atomics International, that operated in Piqua, Ohio, about 34 miles north of Dayton, Ohio. The 

plant contained a 45.5-megawatt (thermal) organically cooled and moderated nuclear reactor that was 

originally built as a demonstration project by the AEC. The reactor was initially operated by Atomics 

International, who also trained employees from the City of Piqua to operate and maintain the plant. 

The City of Piqua eventually took over operation of the reactor, which operated between 1963 and 

1966. 

The primary objective of the POMR facility was to demonstrate the feasibility of the organic reactor 

concept when operated as an integral part of a power generation system, and to provide information 

which ultimately was expected to lead to the design and construction of larger, more economically 

competitive nuclear power facilities (DOE, 2009). The plant was designed as a load-following 

system, meaning the reactor power level varied according to the steam demands. Superheated steam 

at constant pressure was supplied to the steam header in the Piqua Municipal Power Plant, located to 

the west across the Great Miami River. A summary of the POMR operational history is shown in 

Table 5-1. 
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 Table 5-1: Summary of POMR Operational Period History 

Date Action 

June 1963 Initially criticality achieved. 

July 1963 Fuel loading completed. 

January 27, 1964 Full power achieved; reactor operated steadily but with one scram. 

May 21-August 11, 1964 First scheduled shutdown for routine maintenance and inspection 

During this period of operation, POMR contributed ~ 40% of the energy generated by the 

City of Piqua. 

December 7-14, 1964 Reactor was shut down to renew fifteen in-vessel filters and remove the fuel element in 

Core position F-13 for examination. 

January 28-February 14, 

1965 

Reactor was shut down for complete replacement of in-vessel filters, maintenance, and for 

relocation of the instrumented fuel element from position E-12 to position D-5. 

March 12-20, 1965 Planned shutdown to perform physics parameter tests. 

April 2-26, 1965 Several malfunctioning control rod drive units repaired. 

Concern over possible plugged condition of the inner process tube of the control rodbearing elements led to the movement of the six inner ring control rod elements to 

peripheral positions. The core size was increased from 61- 67 fuel elements. 

May 6-12, 1965 Scram occurred on May 6, 1965. During this time, the reactor coolant level had been 

lowered by operational error, which resulted in a temporary loss of circulation through 

three elements. 

Shutdown was extended until May 12 so the three fuel elements could be removed to 

spent-fuel storage. 

May 13, 1965 (estimated 

date) 

Immediately upon restart, excessive surface temperatures were noted, necessitating 

additional fuel element removal. 

Because of the fuel element removal, the system operated with only one coolant pump 

during the latter half of June and into July. 

May 21-25, 1965 Planned shutdown to perform physics parameter tests. 

July 18, 1965 Reactor shut down for modifications, maintenance, and in-vessel filter replacement; 

performed extensive modifications of the in-core control rod circuitry. 

September 6, 1965 Reactor operation resumed. 

October 12, 1965 Reactor shut down, fuel rearrangements were made, increasing the core loading to 70 fuel 

elements. 

October 23, 1965 Reactor restarted. Operation of the reactor continued at an average power level of about 

24 MWt. 

January 13, 1966 Reactor scrammed because of a spurious signal. At this time, there was no indication of 

any unusual condition in the reactor core. Prior to restarting the reactor, an abnormal incore condition was identified during the performance of a rod-drop test. 

Note: The reactor was shut down sometime after the abnormal in-core condition.

Notes: 

This table was created from information included in Summary of Operating History (Unknown author, unknown date-a), 

Piqua Nuclear Power Facility Surveillance and Recovery Program Final Report (Atomics International, 1968b), and from 

information included in Piqua Recovery Program—Phase II-A Plan of Action (Morgan, 1966). 

Three pipelines, approximately 1,400 feet in length, connected the POMR facility on the east side of 

the Great Miami River and the Municipal Power Plant on the west side of the river. These three lines 

included: (1) a 12-inch diameter steam line for steam flow from the POMR steam generator to the 

conventional power plant steam header. This line also provided preheat and start-up steam from the 

conventional plant to the POMR facility, (2) a 6-inch diameter line for boiler feed-water supply from 

the conventional plant to the POMR facility, and (3) a 3-inch line for the return of process steam 

condensate from the POMR facility to the conventional plant. 
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In January 1966, when the reactor shut down due to a scram, several control rod drive problems were 

discovered; in particular, control rod 10 would not drop into place (Atomics International, 1966b). 

The reactor was opened and repairs were made to the control rods. In February, the investigation 

continued; some fuel elements were pulled (or were attempted to be pulled) in an effort to return them 

to their correct positions. However, many of the fuel elements failed to seat properly. On February 4, 

1966, one fuel element was removed and transferred to the fuel storage pool. This fuel element was 

found, using underwater television and photographs, to have buckles at three elevations and to have a 

“carbonaceous material” and a “fouling film” on its surface (Atomics International, 1966b). On 

February 24, 1966, after this fuel element was sent offsite for destructive analysis, the POMR facility 

suspended onsite recovery efforts and began to develop a plan for core unloading (Atomics 

International, 1966b). After discovering the buckled fuel element, operations were secured (through 

May 1, 1966) while the situation was evaluated by Atomics International and a plan for a path forward 

was developed. It should be noted that the buckling that was reported was observed in the outer 

process tube that formed the outer support surface of the fuel element and enclosed the fuel 

assemblies. It is generally believed that most, if not all, of the observed buckles resulted from the 

mechanical forces of the remote handling that was necessary during the lead tube removal from these 

fuel elements, subsequent to their removal from the reactor (Huntsinger, 1969, p. 258). A detailed 

report of the visual inspection of each fuel element included descriptions of each buckle and the 

surface appearance and location of the carbonaceous material (McCurnin, 1966). Nowhere in the 

documents reviewed was there a reference to a fuel cladding failure. 

POMR Components 

Reactor Vessel 

The reactor vessel was a pressure vessel made of low-carbon steel (SA-212-B) and had an internal 

diameter of 7.6 feet, an overall height of 27 feet, and an average wall thickness of 2 inches. It was 

designed to meet applicable American Society of Mechanical Engineers’ codes for an internal 

pressure of 300 psia and 750°F. The vessel contained ten coolant inlet and outlet nozzle penetrations 

that were welded and flanged to the vessel. The penetrations consisted of a single 20-inch coolant 

inlet nozzle, two 14-inch coolant outlet nozzles, a single 6-inch auxiliary inlet nozzle, a 6-inch 

auxiliary outlet nozzle, two 14-inch sampling valve nozzles, two 8-inch control rod cable nozzles, and 

one 8-inch thermocouple lead-through nozzle (Atomics International, 1965a, p. 5). 

Reactor Core 

The core was positioned near the bottom of the reactor core tank. It consisted of a maximum of 85 

fuel elements (the number of fuel elements varied during operations) of slightly enriched uranium and 

13 control-safety rods and was surrounded by an annular thermal shield supported from a ledge inside 

the tank. Steel grid plates, located above and below the core, supported the fuel elements and control 

rods. 

The core pressure was maintained by a control valve in the line to the degasifier. Regulated 

pressurizing pumps returned a constant flow from the degasifier into the primary loop. One 

pressurizing pump had to be running at all times; the second pump could be removed for service. In 

the event of loss of power to the main coolant pump, the pressurizing pumps maintained circulation 

through the core (Atomics International, 1965a, p. 8). 
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Shielding

Reactor radial shielding was achieved through the following: (1) inner thermal shield with 1.5 inch 

steel; (2) outer thermal shield with 4.0 inch steel; (3) reactor vessel wall with 1.125 inch steel; and (4) 

biological shield with 8 feet 4 inches of ordinary concrete (Atomics International, 1965a). 

Shielding at the bottom of the reactor was achieved through the following: (1) the lower grid plate 

with 6-inch steel; (2) organic coolant, measuring 2.5 feet; (3) reactor vessel lower head with 1.125 

inch steel; and (4) vessel support with 3 feet of concrete resting on the reactor building lower-floor 

level (Atomics International, 1965a). 

Shielding above the reactor was achieved through the following: (1) the upper grid plate with 8 inch 

steel; (2) organic coolant, measuring 17.5 feet; and (3) reactor vessel head with 8.5 inch steel 

(Atomics International, 1965a). 

Shielding was provided in other areas that contained process fluids; these areas included the 

purification room, the drain tank room, the waste fired boiler room, and the decay tank room (Atomics 

International, 1965a). 

Control Rods 

The control rod drives consisted of compact unitized assemblies located inside the core tank and 

immersed in the coolant above the core. The neutron absorber element consisted of an assembly of 

outside-diameter tubes filled with boron carbide, positioned to operate inside 13 selected circular fuel 

elements. The drive mechanism operated on the magnetic jack principle, wherein the rod is raised or 

lowered in discrete steps by energizing appropriate sets of electromagnetic coils. 

Fuel Elements 

The fuel elements were circular in cross-section with an outside diameter of 5.25 inches and a length 

of approximately 80 inches (Unknown author, unknown date-b). The uranium inside the elements 

was in the form of two concentric tubes. To improve heat transfer, the inner and outer surfaces of 

each fuel tube were covered with finned aluminum cladding. To maintain physical separation of the 

fuel cylinders, the fins of the cladding were twisted into a slight spiral shape along their longitudinal 

axis. The finned fuel tubes were enclosed between two concentric stainless steel tubes. The ends of 

the steel tubes were fastened to upper and lower end pieces of the fuel elements. The upper end piece 

fit into the upper grid plate and supported the weight of the fuel element. The lower end piece guided 

the element in the lower grid plate and aligned it within the core. 

Fuel elements consisted of uranium fuel, clad with aluminum. The fuel material was a metallic 

uranium alloy, enriched to approximately 1.9 weight percent uranium-235. The alloy was composed 

of uranium, approximately 3.5 weight-percent molybdenum and 0.1 weight-percent aluminum. The 

aluminum cladding had a finned surface to provide an extended heat transfer area. It was 

metallurgically bonded to the uranium fuel using a diffusion barrier of nickel approximately 0.001 

inches thick. The maximum total mass of fuel in the core was 134 kg of uranium-235 and 6,776 kg of 

uranium-238. Figure 5-1 shows a Piqua fuel element. 

An orifice was located in the inlet (upper) end of each fuel element that did not have a control rod 

associated with it. The orifices were adjustable during shutdown and were set to equalize the 

temperature rise across each fuel element. 
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Figure 5-1: Piqua Core I Fuel Element 

Source: Irradiation Performance of Piqua Core 1 Fuel Elements (Arnold, 1966, p. 13) 

Reactor Organic Coolant 

The organic coolant used in the POMR was a commercially available hydrocarbon mixture of the 

three isomers of terphenyl3. After extensive testing of many hydrocarbons, this hydrocarbon mixture 

was selected as the moderator-coolant because terphenyls exhibit relatively high thermal and radiation 

stability, are noncorrosive, and have a relatively low vapor pressure at their expected operating 

temperatures (Unknown author, unknown date-b). This material also had a of about 278 degrees F, 

below which it solidified into a wax-like material (Atomics International, 1961, p. 62). This provided 

an incidental safety feature, as any leak would encapsulate its contaminants somewhat as it cooled to 

room temperature. 

The organic coolant entered the tank above the core and flowed downward through the fuel elements 

into the lower plenum, below the lower grid plate. The coolant then flowed upward through the 

annulus between the core tank and the thermal shield into the outlet plenum of the core tank and back 

to the primary coolant loop. The organic coolant filled all of the available space in the tank and 

served as a moderator, coolant, reflector, and shield for the core. 

During full-power operation, the coolant was heated in the core from 519°F to 575°F while 

transferring 155 x 106

 Btu/hr from the core. The coolant was pumped by two main coolant pumps to 

the super heater and steam boiler where the heat was transferred to the steam system. A total of 

150,000 lb/hr of superheated steam was produced at a pressure of 450 psia and a temperature of 

550°F. The main coolant system consisted of a single loop in which two 6,000 gpm pumps operated 

in parallel, pumping 12,000 gpm to a single super heater and boiler. A flow bypass was utilized to 

divert coolant around the boiler for control purposes. 

 

3

 Terphenyls are aromatic hydrocarbons consisting of three benzene rings linked together with covalent bonds. Terphenyl 

is produced by a process that involves heating the benzene to about 600°C in the presence of a catalyst. 
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Radiation damage to the coolant, as well as exposure to heat, resulted in the formation of various 

gases and low molecular weight compounds, and at the same time produced some higher molecular 

weight compounds (referred to as high boilers). The higher molecular weight compounds consisted 

primarily of long-chained polymerization products and were removed from the coolant by vacuum 

distillation. The high boiler content was permitted to build up to about 30% in the coolant, which was 

thought to be the optimum concentration. The optimization was based on a balance of the decrease in 

radiation and thermal damage (decomposition) rate, and the decrease in the heat transfer 

characteristics of the coolant, as the high boiler content was increased (Unknown author, unknown 

date-b). 

Reactor Systems 

There were a number of auxiliary or supporting systems associated with the operation of the POMR 

facility. Three examples of auxiliary functions included (1) pressurization of the main coolant loop, 

including the reactor vessel, (2) dissipation of the residual or "decay" heat from the reactor core in the 

event the main coolant loop was inoperative, and (3) removal of water, gases, and other low-boiling 

materials formed in the core as a result of coolant decomposition. The schematics associated with the 

POMR systems are relayed in Final Safeguards Summary Report for the Piqua Nuclear Power 

Facility, Appendix K (Atomics International, 1961, pp. 59, 65, 69, 72, 76). 

Containment Building 

The reactor was housed within a containment building that consisted of a steel shell with an inner 

lining of concrete. The steel shell was approximately 73 feet in inside diameter and 123 feet in overall 

height. The shell was fabricated from A-201B steel and was 3/8 inch thick. Thicker plates were 

installed at the various penetration points. The shell was lined with approximately 18 inches of 

concrete on the vertical sections above grade, and the dome was lined with concrete with a varying 

thickness from 18 inches at the lower edge to 6 inches at the top of the dome. The overall free volume 

within the building was approximately 300,000 cubic feet (Atomics International, 1965a, p. 3). 

The containment building was designed to withstand an internal pressure of 5 psig. At this pressure, 

the maximum leakage rate was 1% of the free volume per day. Atmospheric pressure was normally 

maintained within the building by means of supply and exhaust fans (Atomics International, 1965a, p. 

3). Airlocks were provided for access to the containment building. 

Fuel Handling System 

As described in the process and procedures defined in the Final Safeguards Summary Report for the 

Piqua Nuclear Power Facility, Appendix K, the specific fuel handling process was a remote process 

(Atomics International, 1961, pp. 302-305). The spent fuel rods, replaced during refueling operations, 

were transferred remotely to the fuel storage pool. The fuel storage pool was a separate storage 

system that contained an underwater rack for storing the spent fuel rods. Refueling was accomplished 

by working through the reactor top rotating shield, which contained a shield mounted on a circular 

bearing (to permit rotating around the vessel head flange) and a fuel removal port used during 

refueling operations. The fuel removal port could be located over any position in the reactor core. 

The fuel handling cask, which was a shielding device for transferring fuel into or out of the reactor, 

was equipped with a grappling and hoisting mechanism, a traveling bridge and carriage, and an 

emergency cooling system. Also involved in the fuel handling system was a fuel storage pool; the 

fuel storage pool was for the storage of spent fuel elements and consisted of a rack located at the 
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bottom of the pool water. The storage holes in the rack were spaced to prevent fuel element 

criticalities (Atomics International, 1963). 

Heat Transfer System

The main heat transfer system consisted of a single coolant loop which contained two main coolant 

pumps, a superheater, a boiler, a surge tank, and the reactor vessel. Coolant flowed from the reactor 

vessel through the parallel connected main pumps, to the superheater and boiler, and back to the 

reactor vessel. The nominal flow rate through the system with a single pump in operation was about 

7,000 gpm, and with two pumps it was about 12,000 gpm. The flow rate was not less 220 gpm per 

megawatt of thermal power. The heat generated in the reactor core was removed by the circulation of 

the coolant in one downward pass through the core (Atomics International, 1965a, p. 5). 

Degasification System 

The degasification system was designed to remove waste gases and 8 pounds of water vapor per hour 

from the coolant. This system consisted of a degasifier tank which was designed to operate at a partial 

vacuum. The associated piping was provided to supply a sidestream of coolant from the main heat 

transfer system to the degasifier. There was a pressure reducing valve, located upstream of the 

degasifier, that controlled the main heat transfer system pressure and provided accessory control for 

maintaining vacuum in the degasifier. 

Purification System 

The radiolytic decomposition of the coolant produced high-molecular-weight compounds (known as 

high boilers) which had to be removed from the coolant to maintain a fixed percentage of high boiler 

content. The coolant purification system continuously removed these compounds (which were 

subsequently transferred and treated/burned in the waste disposal system); thus, maintaining the 

desired high boiler content in the coolant in the main heat transfer system. The purification system 

also decontaminated the coolant, since most of the radioactivity in the coolant was removed with the 

high boilers (Atomics International, 1961, pp. 66-331). 

As described in Final Safeguards Summary Report for the Piqua Nuclear Power Facility, the 

purification system consisted of a distillation column, column feed heaters, condensers and still 

bottoms, and product receiver tanks. Purification system piping provided a sidestream of coolant 

from the coolant storage system and pressurization system into the column feed heaters. The 

purification system separated high boilers from the coolant and maintained the main heat transfer 

system coolant at a high boiler concentration of approximately 30% or less. The purification system 

was designed to process coolant at a flow rate of up to 1,000 pounds per hour. 

Aqueous Waste System 

The aqueous waste system consisted of a settling basin, waste holdup tanks, demineralizers, and 

interconnecting piping and sumps serving various process areas. The system was designed to gather 

and separate organic and particulate material from the aqueous wastes. Discharge from the settling 

basin could be stored in holdup tanks prior to disposal. Provisions were made for decontaminating the 

water, using an ion exchange technique, if radioactivity levels were above permissible levels. 

Waste Gas System 

The waste gas system consisted of a steam ejector, condensers, and holdup decay tanks. The system 

allowed gases to flow from the purification and degasification systems. The waste gases were 
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processed through one of two decay tank banks; each bank contained eleven decay tanks, each having 

a diameter of 10 inches and a length of 15 feet. Four of the decay tanks in each bank were filled with 

activated carbon for processing waste gas prior to the gas being discharged through the stack. The 

decay tanks had a sufficient total capacity to delay the gases for 48 hours, giving ample time for 

radioactive decay of the process gases before exhausting into the atmosphere. The waste gas system 

provided storage and monitoring of all process gases prior to their release into the atmosphere through 

the stack. 

Organic Waste Disposal System 

The organic waste disposal system consisted of holdup tanks serving the purification system and a 

waste fired boiler designed to burn organic waste from the plant. Still bottoms from the purification 

system column were pumped into the decay tank and stored. The decay tank consisted of a 

compartmentalized vessel with seven compartments, each compartment having a capacity of 3,000 

gallons. The contents of each compartment were sampled prior to processing through the waste fired 

boiler. Organic wastes were burned subsequent to analysis. 

Heating and Cooling System 

The heating and ventilation system provided for air circulation and heating or cooling in all buildings. 

It also provided for maintaining pressure differentials between various areas of the buildings, to 

ensure that air flow was always from non-contaminated areas into those areas having a higher 

potential for contamination. All ventilation air was filtered and monitored before being exhausted into 

the atmosphere. In the event of high radioactivity in the exhaust system, the reactor building had the 

ability to be automatically isolated, and the air would have been recirculated in a closed loop within 

the building. 

POMR Safety Features 

The POMR included many safety features that helped to ensure safe operation of the reactor and 

POMR personnel safety. Notable safety features included concrete for shielding, below-ground level 

components, process gases and waste monitoring, and automatic shutdown and air recirculation. 

Specific safety features include the following: 

 Concrete surrounded the reactor vessel (8’4” of concrete shielded or absorbed radiation emitted 

from the reactor core); 

 Continuous stack effluent monitors (including a particulate monitor and a gaseous activity 

monitor. Detection of airborne radioactivity levels above specified instrument radiation levels 

activated an alarm in the control room, and reactor building isolation devices would be activated 

upon detection of particulate activity);

 Personnel Monitors (Portal Monitor); 

 Exhaust Gas prefilter and absolute filter before going out of the 125 foot exhaust stack; 

 Below-ground-level components (including the core, associated piping, and organic auxiliary 

systems); 

 Solid concrete walls and partitions in the auxiliary building (to protect the environs from radiation 

from the piping and vessels in the building); 

 Sealed openings surrounding the reactor; 

 Containment shell (dome) was maintained at a negative pressure (with respect to the atmosphere) 

and retained any radioactive material released from the reactor vessel or process piping; 
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 Airlocks used for containment shell 

 Noncorrosive properties of the organic coolant (corrosion of the fuel, piping, or other reactor 

equipment was unlikely); 

 Process gases were monitored (prior to entry into the train of decay tanks); 

 Potentially contaminated waste water was stored and monitored for safety determination (prior to 

being discharged from the plant); 

 Radiation monitors located at the exhaust stack (if excessive radioactivity was detected, the reactor 

building was automatically isolated and the air was recirculated within the building); 

 Automatic shutdown actions incorporated throughout the plant (in the event of off-normal 

conditions, the reactor would be shut down and/or isolated); 

 Fifteen remote area monitors (each detector operated a corresponding relay meter and a recorder in 

the control room to continuously indicate the radiation level from the area or equipment being 

monitored); 

 Continuous monitoring of cooling water (via sample stream from the cooling water effluent taken 

prior to the effluent entering and mixing with industrial wastes); 

 Failed element location system (helped determine the presence of a failed fuel element in the core 

by monitoring the delayed neutron activity of the bulk outlet coolant);

 Annunciator point system (provided audio and visual indication of off-normal conditions);

 Fuel handling system (used cameras and periscopes to remotely view the operational areas); and

 Three continuous air monitors (CAMs) were located in the plant

5.1.2 POMR Plant and Process Descriptions for the Post-Operational Period 

Operations were secured in May 1966, due to technical difficulties such as control-rod problems and 

fouling of heat-transfer surfaces. Recovery efforts (attempts to rectify the problems) that were 

intended allow the reactor plant to restart continued into 1967. In December 1967, the AEC 

terminated its contract for reactor operation with the City of Piqua. The Piqua reactor was dismantled 

between December 1967 and February 1969, and the radioactive coolant and most other radioactive 

materials were removed. The remaining radioactive structural components of the reactor were 

entombed in the reactor vessel under sand and concrete (DOE, 2009). A summary of the POMR postoperational history is shown in Table 5-2. 
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Table 5-2: Summary of POMR Post-Operational Period History 

Table 5-2 and its related notes span two pages. 

Date Action 

May 2 - June 6, 1966 Recovery Phase I operations commenced; carried out fuel unloading operations, gridplate measurements, fuel element removal-force measurements, and core position 

“profiling” measurements for carbonaceous material mapping. 

Site investigations revealed the presence of a massive “coke-like” deposit in the outer 

moderator region. 

July 11, 1966 Completed Recovery Phase I operations, which included the following: 

 Removing fuel elements, dummy fuel elements, control rods, source elements, 

Charpy elements, and control rod jumpers from the reactor vessel; 

 Installing support elements in core positions where profiling indicated the presence of 

carbonaceous material; 

 Installing plugs in all lower core grid-plate positions; and 

 Determining the extent of coke formation by profiling the core and organic 

circulation through the reactor vessel.

July 20, 1966 Phase II began, and encompassed the following: 

 Opening the reactor; 

 Replacing the coolant with HB-40 (the Monsanto Chemical Company trade name for 

a hydrogenated-terphenyl mixture); 

 Building a work platform within the outer barrel of the reactor; 

 Removing the carbonaceous material from the reactor; 

 Cleaning the reactor; and 

 Redesigning and modifying the fuel elements. 

There were increased radiation safety requirements due to the unusual nature of the work. 

August 8, 1966 Inner support barrel and upper grid plate were removed from the core and transferred to 

the fuel storage pool. 

 ~50% of coke mass was in an upright position; the remainder had broken off and fallen 

into the lower core-grid plate. 

August 28, 1966 Completed the physical removal of carbonaceous material and sent the material to 

Atomics International for analysis and study. 

September 1966 Circulated (at maximum flow) the hot OMP terphenyl coolant to dislodge and remove any 

particulates remaining in the reactor vessel and system. 

October 1966 Cleaned filtration screens due to accumulation of coke. 

December 31, 1966 The bulk (large pieces that were easily seen) of the carbonaceous material had been 

removed; flushing operations continued into early 1967. 

March 28, 1967 Removed the vessel lid and set the top rotating shield in place. 

Created a temporary ventilation system on the top rotating shield to provide additional 

cooling and to remove fumes from the vessel. The system temperature was about 105 

degrees. The reactor vessel internals could not be inspected immediately because of 

organic fuming. 

July 1967 Filled the reactor vessel and pressurizing and degasification systems with HB-40, which 

was circulated through the reactor and system periodically by each operating crew (about 

1 hour per shift). 

September 1967 Recovery Phase II program activities that were in progress included the following: 

 Flushing the reactor vessel and primary loop piping; 

 Pressurizing the loop piping; and 

 Designing and fabricating modified core components for the restart core. 

October 1967 Terminated the interim flush, the system was cooled and opened; the flushing components 

were removed from the vessel. 
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Table 5-2: Summary of POMR Post-Operational Period History 

Table 5-2 and its related notes span two pages. 

Date Action 

Installed special piping and reestablished circulation. Periodic coolant samples were 

taken during each flush until the filtration and purification systems had reduced the 

coolant impurities to a pre-specified criteria level. 

November 1967 Completed the main heat transfer and degasifier systems. The coolant was drained from 

the system and the reactor vessel was inspected. No material was found in the vessel. All 

core components (fuel, control rods, dummy fuel elements, filters, etc) were removed. 

December 1967 AEC notified the City of Piqua that the plant would be retired. 

May 10, 1968 City of Piqua accepts AEC proposal for “Final PNPF Site Condition.” 

February 1969 With help from several Atomics International employees, the City of Piqua employees 

completed dismantlement activities. The operating authorization was subsequently 

cancelled by the AEC. 

The City of Piqua accepted the plant, and the title to the land occupied by the reactor and 

auxiliary buildings was transferred to the U.S. Government. The site and buildings were 

then leased to the City of Piqua. 

Notes: 

This table was created from information included in Piqua Nuclear Power Facility Surveillance and Recovery Program 

Final Report (Atomics International, 1968b) and Piqua Nuclear Power Facility Monthly Operating Report No. 53

(Atomics International, 1967a).

5.1.2.1 Recovery Period

The period between May 2, 1966 and November 1967 was known as the recovery period. The initial 

plan to return the Piqua plant to normal working conditions consisted of three phases, namely: (1) core 

unloading and inspection, (2) reactor vessel clean-up and core modification, and (3) refueling, testing 

and return to operation (not implemented). At the time of the recovery efforts, the intent was to return 

the reactor plant to full operations. 

The recovery activities were similar to the operations activities; however, because of the coke removal 

and methods to remove the coke, there was increased potential for exposure through non-routine 

conditions (as compared to the operational period). Occurring several times throughout the recovery 

period, the reactor was opened; the fuel elements and the internal support structures were removed and 

stored in a storage pool, then replaced in the reactor; the coolant was flushed through the internal 

systems, and drained and replaced; and filters were changed. While NIOSH is aware that these 

potential exposures were recognized and planned for through AEC requirements (AEC, 1966b), 

NIOSH does not know whether these requirements were met, based on the work that was to be 

performed. 

Phase I: Core Unloading and Inspection 

The core unloading and fuel inspection started May 2, 1966. During this time, radiation 

measurements and core probing took place. The probing revealed the presence of a large 

carbonaceous deposit in the outer moderator region of the reactor that extended over the entire central 

section of the core, bound by the outer control rods. The deposit was not attached to the upper or 

lower grid plates, but was suspended between fuel elements. The volume of the carbonaceous 

material was approximately 4 cubic feet. The fuel elements were visually inspected using closed 

circuit TV and a periscope. After inspection, the fuel elements were removed and transferred to the 
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fuel storage facility. Six of the fuel elements were shipped to the Atomics International Hot Lab 

Facility for further inspection. Recovery Phase I operations were completed on July 11, 1966. 

Phase I accomplished the following tasks (Atomics International, 1968b): 

 Fuel elements, dummy fuel elements, control rods, source elements, and Charpy elements were 

removed from the reactor vessel, inspected, and stored; 

 Control rod jumpers were removed from the reactor vessel; 

 Support elements were installed in the core position where profiling indicated the presence of 

carbonaceous material; 

 Plugs were installed in all lower core grid plate positions; 

 Extent of coke formation was determined by profiling the core; and 

 Organic circulation through the reactor vessel was maintained at approximately 350 degrees. 

Phase II: Reactor Vessel Clean-up and Core Modification 

The reactor vessel clean-up and core modification started on July 20, 1966 and continued through 

November 1967 (Atomics International, 1966d; Atomics International, 1968b, p. 52). The purpose of 

Phase II was to clean the internal reactor components (Atomics International, 1966c). 

Phase II accomplished the following tasks (Atomics International, unknown date-b, p. 2): 

 Carbonaceous material was removed from the reactor vessel; 

 Reactor system was cleaned; 

 Reactor vessel, the system piping low points at the boiler, and three organic holding tanks were 

inspected; and 

 Reactor vessel was modified to provide additional support systems. 

The reactor vessel clean-up required the removal of the upper core-grid plate and barrel supports to 

allow unobstructed access for the mechanical removal (to reduce exposure levels) of the carbonaceous 

material. These two components were removed and replaced multiple times during the cleanup 

activities. While removed from the reactor, both were transferred to the fuel pool for storage. 

The physical removal of the carbonaceous material began August 11, 1966 and was completed on 

August 28, 1966. Much of the larger pieces of the carbonaceous material that were removed were 

shipped to Atomics International for further analysis and study. The smaller pieces, which could not 

be removed mechanically (e.g., using a manipulator), were picked up by an industrial vacuum cleaner 

and were sent to the laboratory to determine the characteristics of the material. 

After the carbonaceous material was removed and sampled, the organic coolant in the reactor was 

replaced with HB-40 because it had a lower freezing temperature than the coolant and allowed a lower 

working temperature while still providing shielding. A work platform was installed inside the outer 

core barrel just above the surface of the HB-40. Air-powered drilling equipment was lowered into the 

vessel and three 3⁄4 inch diameter vent holes were drilled in the outer barrel. The purpose of these 

vents was to eliminate possible trapped gas accumulation and provide local coolant flow. The vent 

holes in the inner core barrel were sealed. At the completion of the drilling operation, the work 

platform and equipment were removed. The ball turret assembly was installed in the top-rotating 
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shield to provide shielding and accommodate cleaning tools, a periscope, and lights. The upper grid 

plate was again removed and the HB-40 was drained from the reactor, and left over coke was 

removed. 

The reactor was refilled with HB-40, which was circulated and filtered to remove any remaining 

pieces of the carbonaceous material. The work platform was used a second time, after again removing 

the inner core barrel and upper core-grid plate, to drill a hole in the barrel support ring and to install 

tubing for new pressure-drop instrumentation. The system was drained, inspected, and refilled with 

purified organic coolant. Flushing and filtering operations for the reactor continued until October 

1967 (Atomics International, 1966e, pp. 9-14; Huntsinger, 1969, pp. 50-52; Atomics International, 

1968b, p. 23). 

The flushing of the main heat transfer and degasifier systems was completed in November 1967. The 

coolant was drained from the system and the reactor vessel was inspected for any remaining 

carbonaceous material. No carbonaceous material was found in the vessel, and the vessel was closed. 

The reactor and primary loop were pressurized to 5 psig with nitrogen to maintain an inert atmosphere 

and to prevent corrosion. All of the core components (fuel, control rods, dummy fuel elements, filters, 

etc.) were removed from the vessel. The coolant was transferred to the high-boiler decay tank, and 

then disposed of by burning in the waste-fired boiler. Only the beryllium sleeve and Charpy impact 

specimen holders remained in the reactor. 

In accordance with instructions from the AEC, the plant was being maintained in a standby condition 

until disposition was agreed upon by the AEC and the City of Piqua (Atomics International, 1968b, 

pp. 51-52). 

Phase III: Refueling, Testing, and Return to Operation 

The goal of Phase III was to load unirradiated Core 1 fuel elements, perform startup testing, and 

achieve power operation. A second part of Phase III included determining the reusability of the fuel 

elements and then reloading acceptable partially-irradiated fuel elements to increase power and extend 

core life (Atomics International, 1966c). However, Phase III was never implemented due to the AEC 

decision to retire the reactor. 

5.1.2.2 Decontamination and Decommissioning

The period between December 1967 and February 1969 was known as the decontamination and 

decommissioning period. A majority of the documents relating to decontamination and 

decommissioning that are in NIOSH’s possession are planning documents; there are no after-action 

reports indicating how the operations actually proceeded with respect to health and safety measures. 

A summary of activities that were performed during the decontamination and decommissioning period 

include the following: 

 Cutting and preparing contaminated piping for shipping and disposal; 

 Using mechanical and flame-cutting methods; 

 Welding shut reactor openings; 

 Manipulating and shipping contaminated structures and irradiated fuel assemblies; and 
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In December 1967, the AEC notified the City of Piqua that the plant would be retired and that they 

would terminate the contract for the operations and maintenance of the POMR. At this time, 

arrangements were initiated to remove the fuel elements and many of the reactor’s internal 

components from the site, and to dispose of the organic coolant. After several case studies, it was 

concluded that the process of removing the reactor vessel would constitute a greater potential hazard 

than not removing the vessel, so the decision was made to leave the reactor vessel in place. 

The AEC wrote the proposal for the “Final PNPF Site Condition” on March 27, 1968 and submitted it 

to the City of Piqua on April 28, 1968. The proposal described the general requirements for the site at 

the end of dismantlement and was included as part of the retirement plan. The City of Piqua accepted 

the proposal on May 10, 1968 (Atomics International, 1968d, p. 26). 

Early in the preliminary retirement planning stages, the AEC decided to retain overall responsibility 

for the management and administration of the dismantlement effort. The City of Piqua was 

responsible for supervising and conducting the on-site work, mostly with the remaining operating 

crew, but also with the assistance of outside contractors who were retained for work at the option of 

Piqua. Atomics International was responsible for the engineering effort, which included preparation 

of plans, safety analyses and reports, studies of alternatives, engineering for the facility modification, 

and preparation of “as modified” drawings and final reports. Both the AEC and Atomics International 

maintained representatives at the site during most of the dismantlement period. The City of Piqua 

operating crew (approximately 25 people) included the reactor superintendent, several licensed 

operators, health physics personnel, and office help. The City of Piqua personnel directed all field 

work and performed most of the actual material removal (Atomics International, 1968d, pp. 26-28). 

Dismantlement 

When the decision was made to retire the POMR, the plant had not operated in almost two years due 

to modification and repair work. During this two-year period, an essentially complete operating crew 

was employed, and these same personnel were used in the dismantlement effort. Dismantlement 

activities outside of the reactor complex (beyond the first blocking valves) were completed under a 

partial dismantlement authorization, and the remaining work was accomplished after the final order 

was issued. 

The broad scope of the dismantlement was gradually performed over a one-year period. The 

deactivation work was completed area-by-area throughout the facility. This included removing fuel, 

enclosing the reactor vessel in the concrete shielding, removing equipment and material (contaminated 

and uncontaminated) from the reactor and auxiliary building, and restoring the buildings for use as a 

warehouse/office. The reactor vessel, thermal shield, grid plates, and support barrels were left in 

place. The nozzles and piping that penetrated the concrete shield were cut off and welded shut. The 

sleeves through which the piping passed were plugged and a new surface was applied to the concrete 

shield to cover the plugged penetrations. The reactor lid was bolted onto the reactor vessel and tests 

were performed to verify that there were no gross leakages from the reactor vessel. The floor plug 

over the reactor vessel was tack welded into place to prevent access to the reactor vessel top-flange 

(Atomics International, 1968a, pp. 9-10). 
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As of April 1968, the following major activities had been completed: 

 The reactor core components (except two beryllium sleeves and two Charpy impact specimen 

holders) had been removed from the reactor vessel and were stored in the spent fuel storage pool; 

 A total of 76 irradiated fuel elements were being stored in the spent-fuel storage pool; 

 All unused fuel had been removed from the site; and 

 The organic coolant had been drained from all coolant systems into the high-boiler decay tank (T9) and had been burned in accordance with approved procedures. 

In the decommissioning of Piqua, nuclear fuel, coolant, and much of the equipment were removed. 

The reactor vessel, concrete shielding, and many other items containing radioactive material were left 

in place, securely contained in the reactor complex. The contaminated piping and equipment inside 

the reactor building was removed or decontaminated and the reactor complex was closed. The floor 

of the above-ground portion of the reactor building was covered with a waterproof material and a 

layer of concrete so the regions containing the radioactive material would be completely inaccessible. 

A few kilocuries of activation products remain inside the reactor complex within its eight-foot thick 

concrete walls. The radioactivity is an integral part of the reactor structures such as the massive six 

inch thick steel, lower grid plate. Two time capsules, with detailed information on the design and 

content of the reactor complex are located in the reactor building. One time capsule is located on top 

of the reactor vessel and below the top-shield plug. The second time capsule is located in a recess in 

the shield wall near the door into the auxiliary building. 

Final Piqua Report 

Upon completion of the work in the approved dismantlement plan, a report titled Retirement of the 

Piqua Nuclear Power Facility was submitted to the Atomic Energy Commission, Division of Reactor 

Licensing (AEC-DRL), which stated the following: 

 The facilities were decontaminated and dismantled in accordance with the dismantlement order; 

 The protective systems and all other work described in the dismantlement plan were completed; 

 A final inspection had been completed by the AEC Division of Compliance and by an independent 

certified health physicist; 

 There were no findings which would prevent termination of the Piqua Nuclear Power Facility 

Operating Authorization; and 

 Suggested that the Operating Authorization be terminated, in accordance with the dismantlement 

order and applicable AEC regulations. 

Upon completion of the dismantlement program, cancellation of the operating authorization, and 

acceptance of the plant by the City of Piqua, in February 1969, the title to the land occupied by the 

reactor and auxiliary buildings was transferred to the U.S. Government. The site and buildings were 

then leased to the City of Piqua (Wheelock, 1970, p. 47). 

5.2 Internal Radiological Exposure Sources from Operational and PostOperational POMR Activities 

ATTRIBUTION: Section 5.2 and its related subsections were completed by Daniel Mantooth, Dade 

Moeller and Associates, Inc. and Eugene W. Potter, M. H. Chew & Associates, Inc. These 
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conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for all 

conclusions in this document are explained in the associated text. 

The potential sources of alpha-emitting radionuclides at the POMR site included the uranium fuel 

itself, as well as transuranic radionuclides produced during the fission process while the reactor was 

running. However, in either case, these alpha-emitters would not be available for personnel exposures 

unless the aluminum cladding on the fuel was breached, allowing these radionuclides to contaminate 

the coolant and other parts of the system. The pre-demolition radiological survey performed in 1968, 

reported alpha contamination levels as “...non detectable during operation and less than 1 dpm/100 

cm

2

 in final survey in all locations” (Wheelock, 1970, Appendix B). 

The possibility of a cladding breach was discussed in one document (Ashley, 1964, p. 173) as being 

the cause of low levels of xenon-133 (a uranium fission product) detected in the process gas. 

However, the fact that the failed Fuel Element Location System (FELS) did not indicate such a failure 

led POMR site engineers to conclude that the xenon-133 was most likely a result of uranium 

contamination in the aluminum cladding. In any case, the amount of uranium involved would not 

have resulted in measurable exposures to POMR personnel. The absence of alpha emitting 

radionuclides was confirmed by radiological surveys conducted prior to the facility retirement 

(Wheelock, 1970, Appendix B, Table B-4). 

The primary sources of internal exposure at the POMR facility consisted of beta/gamma-emitting 

radionuclides from four sources: (1) activated impurities in the coolant, (2) activated corrosion 

products, (3) neutron recoil reactions with the aluminum cladding, and (4) tritium produced by ternary 

fission.4

Based on NIOSH’s evaluation, significant issues existed in regard to the personnel exposure sources 

and exposure conditions during the recovery and decontamination and decommissioning phase 

activities, as compared to the operational period at the site. Examples of these issues include the 

following: 

 Opening systems for intrusive recovery activities. 

o Removal of fuel elements and interior reactor system fuel element support structures; 

o Access to the open reactor vessel; and 

o Removal of carbonaceous material from reactor systems. 

 Intrusive activities associated with decontamination and decommissioning work. 

o Burning, cutting, and welding reactor systems and components 

 Potentially inadequate administrative controls. 

o Use of CAMs which may have been sufficient to support monitoring during closed system 

operations. However, it is not apparent that the number of CAMS and their “semi-portability” 

would have been adequate to properly characterize workplace activities during the recovery 

period and would be less likely to be adequate during the decontamination and 

decommissioning period; 

o Lack or insufficient use of radiological administrative controls— including local ventilation, 

respiratory protection, or personal protective equipment. 

4

 Attribution: The percentage reported for each source was calculated by Daniel Mantooth, Dade Moeller and Associates, 

Inc. from estimated concentration values found on page V-4 of Atomics International, 1961. 
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5.2.1 Activated Impurities in the Coolant 

The radionuclides that resulted from activation of impurities in the coolant, as specified in Table 5-3, 

included sodium-24, phosphorus-32, sulfur-35, chlorine-38, manganese-56, arsenic-76, argon-41, and 

nitrogen-13 (Atomics International, 1961; Atomics International, 1965a; Atomics International, 

1964a), although argon-41 was believed to result from air in-leakage to the reactor vessel. The 

presence of nitrogen-13 was explained in POMR-related documentation as “...from the nitrogen cover 

gas” (Atomics International, 1965a). However, it is more likely that the nitrogen-13 was created from 

the carbon in the organic coolant from the carbon-12 (p, γ) nitrogen-13 reaction, or the oxygen-16 (p, 

α) nitrogen-13 reaction. The gaseous species argon-41 and nitrogen-13 were removed by the coolant 

purification system, and after sufficient decay were exhausted via the main stack. Personnel had little 

opportunity for exposure to these gases during maintenance activities. Due to their short half-life, the 

majority of the activity in the system would have decayed prior to breaching coolant or other systems. 

POMR personnel outside the facility may have been exposed to low levels of these radionuclides 

during reactor plant operation. 

Ninety-six to ninety-eight percent of the non-gaseous radionuclides sodium-24, phosphorus-32, 

manganese-56, and arsenic-76 were thought to be collected along with the high-boiling organics from 

the coolant purification system. Seventy-eight percent of the sulfur-35 and chlorine-38 were collected 

in the same manner (Atomics International, 1961). Other systems where particulate activity could 

have been encountered included the in-vessel filters, coolant purification system, and the fuel storage 

pool purification filters. Operations that involved system maintenance or filter material replacement 

would have carried the potential for exposing personnel to these radionuclides. These radionuclides 

were present throughout the operational history of the reactor, but due to radioactive decay, the 

relative contributions would shift to longer-lived species during an outage period, or during the postoperational period. Activated impurities were estimated to comprise 48% of the particulate 

radioactive species in the coolant (Atomics International, 1961). 

5.2.2 Activated Corrosion Products 

The primary radionuclides that resulted from the activation of corrosion products in the coolant were 

manganese-54, manganese-56, chromium-51, iron-55, and iron-59. Since these radionuclides are 

particulates, the discussion pertaining to locations in the reactor and mechanisms of exposure are 

identical to that presented for activated coolant impurities in Section 5.2.1, above. Activated 

corrosion products were estimated to comprise 13% of the total particulate activity (Atomics 

International, 1961). The radiological data for the potential sources are listed in Table 5-3. 

The reported method for radionuclide identification in the reactor coolant and effluents was by pulseheight analysis (gamma spectroscopy). This method would not have detected non-gamma emitting 

species such as tritium or carbon-14. The presence of tritium and carbon-14 was indicated in the 

Final Safeguards Summary Report for the Piqua Nuclear Power Facility, with calculated 

concentrations in coolant of 5.3E-04 μCi/cc and 1.4E-04 μCi/cc for tritium and carbon-14, 

respectively (Atomics International, 1961). Analytical results for tritium and carbon-14 in the coolant 

have not been found. Analyses of various process residues and water (Wheelock, 1970 Appendix B, 

Table B-2) in the fuel-storage pool indicated tritium levels ranging from 7.5E-04 μCi/cc to 2.1E-02 

μCi/cc. Results for carbon-14 ranged from 9E-05 μCi/cc to 2.1E-02 μCi/cc. The Piqua Nuclear 

Power Facility Monthly Operating Report No. 24 reported that “bioassay performed for personnel 
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working over the open reactor for tritium and net beta activity showed no positive result” (Atomics 

International, 1965b); these bioassay data have not been located. 

5.2.3 Neutron Recoil Reaction Products 

Several particulate nuclides were thought to be produced by neutron recoil reactions with materials 

present in the aluminum fuel cladding (Atomics International, 1961; Atomics International, 1964c; 

Atomics International, 1965a; Atomics International, 1964a; Atomics International, 1965d). These 

include sodium-24 (also arises from impurities), magnesium-27, cobalt-58, and cobalt-60. [Note: 

Magnesium-27 has a half-life of 9.5 minutes. Isotopes with half-lives less than 10 minutes provide 

negligible contribution to internal dose. Thus, magnesium-27 will not be included as a radionuclide of 

concern.]5

 Since these radionuclides are particulates, the discussion pertaining to locations in the 

reactor and mechanisms of exposure is identical to that previously presented for activated coolant 

impurities in Section 5.2.1. Radionuclides resulting from recoil reactions were estimated to comprise 

39% of the total particulate activity (Atomics International, 1961). 

5.2.4 Fission Products 

Potential fission products such as xenon-133m, xenon -135, krypton-85m, krypton-85, and krypton-87 

were detected in the process gas, but only at levels on the order of a few disintegrations per minute per 

gram (dpm/g). Considering the low activity levels and the fact that these isotopes are noble gases and 

are only significant from an external exposure standpoint, they will not be considered further as a 

source of internal dose. 

The Final Safeguards Summary Report for the Piqua Nuclear Power Facility proposed that the 

presence of tritium in the coolant was primarily a result of the ternary fission in the fuel (Atomics 

International, 1961, p. V-6). Twenty-five percent of the tritium created was assumed to enter the 

coolant by diffusion or recoil processes, which would have resulted in an estimated equilibrium 

concentration of 0.21 μCi/cm3

 at full power. This concentration is 10 times greater than the maximum 

value measured in process residues reported in Section 5.2.2, which were analyses of various process 

residues in systems and water in the fuel-storage pool. Analytical results for tritium in the coolant, 

nor environmental monitoring data have been located. However, an analysis for tritium in the 

carbonaceous material that was discovered in the coolant resulted in a level of 3.9 μCi/g (Atomics 

International, 1966b). The Piqua Nuclear Power Facility Monthly Operating Report No. 24 

mentioned that “bioassay performed for personnel working over the open reactor for tritium and net 

beta activity showed no positive result” (Atomics International, 1965b). However, no bioassay data 

have been located. 

5

 Attribution: This decision was based on comments from the ORAU Team’s DOE Site Principal Internal Dosimetrist, 

Elizabeth Brackett, MJW Corporation. 
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Table 5-3: Radionuclides of Concern for POMR Operational and Post-Operational Internal Exposures

Radionuclide Half-life* Primary Radiations 

and Energies (Mev)* Potential Source Reactions

Na-24 15 hr β-

 4.17 

γ 1.369 

Coolant Impurities: 

Recoil:

23Na(n,γ)24Na* 27Al(n,α)

24Na**

P-32 14.28 d β- 1.71 Coolant Impurities: 32S(n,p)32P*

S-35 87.9 d β- 0.167 Coolant Impurities: 

Coolant Impurities:

37Cl(d,a)35S* 34S(n,γ)

35S*

Cl-38 37.29 min β-

 4.91 

γ 1.60 Coolant Impurities: 37Cl(n,γ)

38Cl*

Cr-51 27.8 d Electron Capture 

V x-rays 0.320 (9%) Corrosion: 50Cr(n,γ)

51Cr*

Mn-54 303 d

Electron Capture 

Cr x-ray 0.835 (100%) 

β-

 0.829

Corrosion: 54Fe(n,p)54Mn*

Mn-56 2.58 hr β-

 2.85 

γ 0.847 Corrosion/Impurities: 55Mn(n,γ)

56Mn*

Fe-55 2.6 yr

Electron Capture 

Mn x-rays 

bremsstrahlung 

continuous to 0.23

Corrosion: 54Fe(n,γ)

55Fe*

Co-58 70.88 d β+

 0.474 

γ 0.810 (99%) Recoil: 58Ni(n, p)58Co**

Fe-59 45.6 d

β-

 1.57, 0.475 

γ 1.095 (56%) 

 1.292 (44%) 

 0.192 (2.8%)

Corrosion: 58Fe(n,γ)

59Fe*

Co-60 5.26 yr

β-

 0.314 (99%) 

γ 1.173 (100%) 

 1.133 (100%)

Recoil: 60Ni (n,p)60Co*

As-76 26.4 hr

β-

 2.97 

γ 0.559 (43%) 

 0.957(6%) 

 1.22 (5%) 

 ≤2.1 (2%)

Coolant Impurities: 75As(n,γ)

76As*

Tritium 12.3 yr β- 0.0186 Coolant Activation 2H(n,γ)

3

H*

C-14 5730 yr β- 0.156 Coolant Activation 13C(n,γ)

14C*

Notes: 

* Data are from Radiological Health Handbook (Radiological Health Handbook, 1970).

** Data are from Final Safeguards Summary Report for the Piqua Nuclear Power Facility (Atomics International, 1961). 

5.3 External Radiological Exposure Sources from Operational and PostOperational POMR Activities 

ATTRIBUTION: Section 5.3 and its related subsections were completed by Louise Buker, Oak Ridge 

Associated Universities (ORAU) and Roger Halsey, Oak Ridge Associated Universities (ORAU). 

These conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for 

all conclusions in this document are explained in the associated text. 

Workers at the POMR facility were potentially exposed to external photon, beta, and/or neutron 

radiation from activities associated with maintaining and operating the reactor. Potential exposure 
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sources included radioactive materials from the operation and maintenance of the nuclear reactor and 

radioactive materials in the form of calibration sources. 

As discussed in Section 5.2, significant issues existed in regard to the personnel exposure sources and 

exposure conditions during the post-operational period (i.e., recovery activities and decontamination 

and decommissioning phase activities) that did not exist during the operational period at the POMR 

site. 

5.3.1 Photon 

Some POMR radiological operations potentially involved gamma and X-ray photon radiation fields. 

The potential photon exposure sources would have included the following: 

 Gamma-emitting fission and/or activation products resulting from the reactor operations, 

 Bremsstrahlung radiation from various beta-emitting radionuclides, and 

 Calibration sources of cobalt, cesium, and other miscellaneous radionuclides (Geiger, 1969). 

During operations, the radiation levels from the coolant lines and filters would have resulted from 

activation of corrosion products and other impurities in the coolant, as well as from recoil products 

from the core. The short lived gamma-emitting isotopes that have a half-life of 10 minutes or less 

(magnesium-27 and nitrogen-13) would be responsible for the majority of the radiation from the 

coolant lines (Atomics International, 1965a, p. 265). During reactor operations, the highest radiation 

levels in the plant were at the degasifier filters, F 2A and F 2B (Atomics International, 1965a, pp. 265 

-270). 

During the post-operational period, there would have been additional potential for exposures to fuel 

assemblies and highly activated internal components (primarily handled remotely, but occasionally 

manipulated with ad hoc methods) (Atomics International, 1966e, p. 14). In addition, AEC expected a 

potential for exposure to uneven fields during the recovery phase, leading to a requirement for wrist 

badges to be used to monitor for extremity exposure, in addition to whole-body badging (AEC, 

1966b). 

5.3.2 Beta 

There was a potential for beta particle-emitting source term during shutdowns, maintenance, refueling, 

and when fuel was removed. Beta radiation could have resulted from activation and fission products. 

During normal operations, fission and activation products would have been located within the core 

and within the various shielding surrounding the core. 

NIOSH located documentation explaining the beta radiation levels during maintenance activities. The 

documentation stated that “beta radiation is the primary type of activity encountered when systems are 

opened.” It further stated that through the entire test program, as a result of the low dose rates present, 

it was unnecessary to establish work time limits for personnel performing maintenance work on 

exposed system components (Atomics International, 1965a, p. 275). 

The potential for beta exposure from fission and activation products was known and documented prior 

to plant startup (Atomics International, 1963, p. 83). 
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During the post-operational period, the potential for exposure would remain very similar to the 

operational period. The isotopes present would, however, vary after shutdown due to their differing 

half-lives. 

5.3.3 Neutron 

There was a low potential for neutron radiation exposure associated with POMR operations. Neutron 

exposures could have occurred, as a result of the fission process, from operating the reactor from 1963 

to 1966. In addition, there was some potential for neutron exposures from a plutonium-beryllium 

(PuBe) neutron source for those personnel who calibrated the neutron survey instruments (Personal 

Communication, 2009g). The potential personnel exposures from this type of check source is 

considered to be encompassed in the assessment of reactor-related neutron exposures included in this 

evaluation. The rate of neutron generation from reactor operations was based on the rate of fission 

which would have been directly proportional to the reactor power level. An interviewee mentioned 

that there was a plutonium-beryllium neutron calibration source used for checking the long counter6

, 

which was an energy-independent neutron measuring instrument (Personal Communication, 2009g). 

The source of the neutron emissions and potential worker exposures was minimized as a result of the 

design of the POMR, which controlled the operationally-related neutron exposure sources. The 

POMR was well-shielded by concrete that surrounded the reactor vessel, below-ground-level 

components, and sealed openings that surrounded the reactor (Unknown author, unknown date-b). 

During the post-operational period, there would be no neutron exposure potential from the reactor. 

6.0 Summary of Available Monitoring Data for the Class Evaluated 

by NIOSH 

The following subsections provide an overview of the state of the available internal and external 

monitoring data for the POMR class under evaluation. 

6.1 Available POMR Internal Monitoring Data 

ATTRIBUTION: Section 6.1 and its related subsections were completed by Daniel Mantooth, Dade 

Moeller and Associates, Inc. and Eugene W. Potter, M. H. Chew & Associates, Inc. These 

conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for all 

conclusions in this document are explained in the associated text. 

Section 6.1 and its related subsections summarize the internal monitoring data for the operational and 

post-operational periods at the POMR facility. 

 

6

 A long counter is a neutron detector that is designed to measure all neutrons accurately. It has a flat response over a 

broad range of neutron energies. At its core, it has a boron trifluoride proportional counter, able to discriminate neutrons 

against gamma activity by use of the neutron-alpha reaction on boron-10. This core is surrounded by a cylinder of 

paraffin. Fast neutrons are moderated by the paraffin and are then captured by the boron-10 in the boron trifluoride gas. 

Discriminators are used to accept only the signals from the alpha and lithium-7 recoil products. The detectors are energy 

independent, typically in the range from 10 keV to 5 MeV and are shielded to provide a highly directional response 

(Cember, 1969). 
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6.1.1 Available POMR Internal Monitoring Data for the Operational Period 

No primary source bioassay monitoring data (e.g., original dosimetry records) were located for POMR 

personnel for the operational period. One report, Piqua Nuclear Power Facility Monthly Operating 

Report No. 24 mentions that urine samples were collected for personnel “...working over the open 

reactor,” but the bioassay showed no positive result tritium and net beta activity. One former 

employee reported that bioassays were performed annually (Personal Communication, 2009h). 

 

Maximum and occasionally average stack effluent activity concentrations were reported in monthly, 

quarterly, and semiannual reports (Atomics International, 1964b; Atomics International, unknown 

date-a; Atomics International, 1965b; Atomics International, 1965d; Atomics International, 1964c). 

The concentrations reported in the reports are consistent with the calculated estimates provided in the 

Final Safeguards Summary Report for the Piqua Nuclear Power Facility (Atomics International, 

1961, Table V-3). 

Maximum and average air activity concentrations for environmental and onsite samples are provided 

in three reports—Piqua Nuclear Power Facility Monthly Operating Report No. 24; Piqua Nuclear 

Power Facility Monthly Operating Report No. 26; and Piqua Nuclear Power Facility Monthly 

Operating Report No. 19. The location of these samples could not be determined from the available 

information. The concentrations reported are substantially less than the maximum permissible 

concentration values shown in Table V-3 of Final Safeguards Summary Report for the Piqua Nuclear 

Power Facility (Atomics International, 1961).

No primary source workplace air or breathing zone air sample data (e.g., original records) were 

located. However, there were instances in which levels were discussed in relative terms. For 

example, a monthly operating report states “...no personal contaminations or inhalations” (Atomics 

International, 1965b), and a progress report states that “Airborne activity in the containment building 

has not exceeded that normally observed from natural background” (Atomics International, unknown 

date-a). 

6.1.2 Available POMR Internal Monitoring Data for the Post-Operational Period 

No primary source bioassay monitoring data (e.g., original dosimetry records) were located for POMR 

personnel for the post-operational period. 

Maximum, minimum, and average air concentrations for stack, on-site, and environmental filters were 

only located in a single monthly report (September 1967) for the entire post-operational period 

(Atomics International, 1967a). 

6.2 Available POMR External Monitoring Data 

ATTRIBUTION: Section 6.2 and its related subsections were completed by Louise Buker, Oak Ridge 

Associated Universities (ORAU) and Roger Halsey, Oak Ridge Associated Universities (ORAU). 

These conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for 

all conclusions in this document are explained in the associated text. 

Section 6.2 and its related subsections summarize the external monitoring data for the operational and 

post-operational periods at the POMR facility. 
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6.2.1 Available POMR External Monitoring Data for the Operational Period 

NIOSH has not located any primary source external dosimetry records (e.g., original dosimetry 

records) for the Piqua workers during the operational period. External data for the POMR facility that 

have been found are in the form of exposure summary reports and monthly and semiannual 

operational progress reports for the years 1963 through 1966. The historical Piqua documentation 

demonstrates knowledge of potential workplace external radiation hazards, applicable radiation 

exposure guidelines, methods for limiting worker exposure, and radiation monitoring and dosimetry 

capabilities. 

Photon

The most complete, available data consist of summary dosimetry data reported annually to the AEC 

offices for Chicago Operations and Idaho Operations. These summary data provide evidence that 

external monitoring was performed for the Piqua site. Table 6-1 summarizes the dosimeter results for 

the operational period that were sent annually to the AEC Chicago Operations Office. 

Table 6-1: Summary of Annual Whole-Body Penetrating Exposures for the POMR Operational Period 

Year No. Identified as 

NOT Monitored 

No. Identified as 

Monitored 

No. Identified with 

0-1 rem 

No. Identified with 

1-2 rem 

1963 0 42 42 - 

1964 0 46 46 - 

1965 0 47 47 - 

1966 0 50 42 8 

Notes: 

- indicates that no exposures were reported in this range. 

Information for 1963 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1963 (AEC, 1963). 

Information for 1964 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1964 (AEC, 1964). 

Information for 1965 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1965 (AEC, 1965a). 

Information for 1966 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1966 (AEC, 1966a). 

The summary data that were reported to the AEC Idaho Operations Office for the year 1963 indicated 

that 31 Atomics International employees worked at the Piqua facility with the City of Piqua workers, 

performing the same types of jobs. The Atomic International employees’ dosimeters indicated that 

they received 0 gamma, 0 beta, and 0 neutron dose for the year 1963 (City of Piqua, 1964). One 

worker did have detectable gamma of 60 mrem for the year. Individual data for the City of Piqua 

workers are unavailable at this time. 

Monthly and semiannual operational reports indicate that R. S. Landauer Jr. and Company was the 

dosimetry provider. However, individual dosimetry records are not currently available; thus, the 

information that NIOSH has available to supplement the data provided in Table 6-1 consists of 

operational progress reports from 1963 through 1966. These reports include references to ambient 

radiation levels and routine survey results. Table 6-2 provides a summary of the available 

operational-period data from semiannual and monthly reports that were sent to the Chicago AEC 

Operations Office. 
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Table 6-2: Semiannual and Monthly Report Results for the POMR Operational Period 

Table 6-2 and its corresponding notes span two pages. 

Timeframe Results 

Report Type 

and Number Begin End 

Number of 

Positive 

Results 

Maximum 

Result 

Notes from the Radiation Monitoring 

Section of the Reports 

Semiannual-2 

(Atomics 

International, 

1963) 

01/01/63 06/30/63 Unknown Unknown No radiation measurement mentioned 

(Atomics International, 1963) 

Semiannual-3 

(Atomics 

International, 

1964d) 

07/01/63 12/31/63 Unknown Unknown No radiation measurement mentioned 

(Atomics International, 1964d) 

Semiannual-4 

(Atomics 

International, 

1964a) 

01/01/64 06/30/64 Unknown Unknown Radiation levels measured at eighteen 

locations while operating at full power (45.5 

MWt.) 

No neutrons detected from the biological 

shield or above the reactor at the 100’ level 

(Atomics International, 1964a, p. 82) 

Location OC-104 <0.5 mrem/hr neutrons 

Location OC-101, 14" line in the P-1A room 

<0.5 mrem/hr neutrons (Atomics 

International, 1964a, p. 83) 

During shutdown in June, the following 

measurements were taken: 

 1 R/hr measured at in-core filter that had 

been removed (Atomics International, 

1964a, p. 17) 

 Filter media measured < 20 mrad/hr beta 

- gamma, including 5 mR/h gamma 

(Atomics International, 1964a, p. 17) 

Semiannual-5 

(Atomics 

International, 

1965e) 

07/01/64 12/31/64 Unknown Unknown 2-4 mR/h surface of main coolant piping 

(Atomics International, 1965e, pp. 61-63) 

Monthly-19 

Atomics 

International, 

1964c) 

10/01/64 10/31/64 10 30 mrem N/A 

Semiannual-6 

(Atomics 

International, 

1965c) 

01/01/65 06/30/65 Unknown 120 mrem During the month of April, this max 

estimated dose was received during 

inspection from non-routine removal of fuel 

elements (Atomics International, 1965c, pp. 

76-77) 

700mR/h degasification filter, attributed to 

Mg-27 and Na-24 from the 6 new fuel 

elements (Atomics International, 1965c, pp. 

76-77) 
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Table 6-2: Semiannual and Monthly Report Results for the POMR Operational Period 

Table 6-2 and its corresponding notes span two pages. 

Timeframe Results 

Report Type 

and Number Begin End 

Number of 

Positive 

Results 

Maximum 

Result 

Notes from the Radiation Monitoring 

Section of the Reports 

Monthly-24 

Atomics 

International, 

1965b) 

03/01/65 03/31/65 4 80 mrem N/A 

Semiannual-7 

(Atomics 

International, 

1966a) 

07/01/65 01/13/66 Unknown Unknown No section for plant radiation levels; some 

discussion of dose rates for a degasifier 

(Atomics International, 1966a, p. 66) 

Monthly-26 

(Atomics 

International, 

1965d) 

05/01/65 05/31/65 2 290 mrem Dose received by HP personnel during 

reactor shutdown monitoring and calibration 

of HP instruments (Atomics International, 

1965d, p. 20) 

Semiannual-8 

(Atomics 

International, 

1966b) 

01/01/66 06/30/66 Unknown Unknown 5 mR/h beta and gamma; sample of coke 

surface (Atomics International, 1966b, p. 

103) 

Notes: 

 “Unknown” indicates that information was not provided or available. 

For 1963, the available operational reports—Piqua Nuclear Power Facility Operations Analysis 

Program Progress Report No. 2 (Fiscal Year 1963) and Piqua Nuclear Power Facility Operations 

Analysis Program Progress Report No. 3 (July 1, 1963-December 31, 1963)—did not discuss the 

highest dose for POMR personnel (Atomics International, 1963; Atomics International, 1964d). 

However, there is a 1963 report from the City of Piqua for 31 Atomics International workers that 

worked at the site; this report lists 0 gamma, 0 beta, and 0 neutron dose for thirty of the employees. 

For one employee, the report lists 60 mrem gamma (City of Piqua, 1964). In addition, there is a 

summary of quarterly cycle data that was submitted to the AEC for one Piqua employee for the years 

1963 through 1965, where it was reported that he received 0 photon, 0 beta, and 0 neutron dose for 

each of those years (AEC, 1965b). 

For 1964, Piqua Nuclear Power Facility Monthly Operating Report No. 19 reported only one 

maximum exposure of 30 mrem for the month of October, and 10 readings that were greater than 

detectable levels for deep dose (Atomics International, 1964c). The semiannual operational reports 

did not mention the highest exposure for the period. The reports include summaries of survey 

monitoring data that provide dose rates at certain areas of the plant for beta, gamma, and neutrons. 

Additionally, Piqua Nuclear Power Facility Operations Analysis Program Progress Report No. 4 

(January 1, 1964-June 30, 1964) provides a summary of the radiation levels associated with the 

megawatt thermal energy of the reactor from January through May of 1964 for survey monitoring 

dose rates at the “waste gas activity” location, the “out of core bypass filter activity” location, and the 

“main coolant superheater outlet line” location. The highest level was approximately 400 mR/hr at 

the “out of core bypass filter” location; the “waste gas activity” and the “main coolant superheater 

outlet line” measurements were less than 30 mR/hr (Atomics International, 1964a, p. 80). 
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For 1965, the available operational reports— Piqua Nuclear Power Facility Reactor Operations 

Analysis Program Semiannual Progress Report No. 6 (January 1-June 30, 1965) and Piqua Nuclear 

Power Facility Monthly Operating Report No. 24—report that there were maximum doses reported of 

80 mrem in March; 120 mrem in April during a non-routine inspection of fuel elements; and 290 

mrem in May, received during calibration activities during a reactor shutdown period (Atomics 

International, 1965c; Atomics International, 1965b). 

Beta

NIOSH has not located a complete set of data covering beta or shallow measurements. The monthly 

and semiannual reports have reported beta or shallow measurements occasionally with gamma 

measurements. The annual summary reports to the AEC Chicago Operations (the whole-body 

penetrating exposures are provided in Table 6-1) do not provide a beta dose because only whole-body 

penetrating exposure ranges were being reported. 

However, the external dosimetry for plant workers (film badges provided by R. S. Landauer Jr. and 

Company) included a shallow dose component (Atomics International, 1965b, p. 23). In the summary 

data that were reported to the AEC Idaho Operations Office for 1963, the Atomic International 

employees’ dosimeters indicated that 30 of the 31 individuals monitored received 0 beta dose for the 

year. One individual had a reported dose of 20 mrem beta (City of Piqua, 1964). 

Survey data reported in the annual and semiannual reports typically were reported as β-γ (betagamma). There were a couple of instances where the gamma exposure rate was distinguished from 

the beta-gamma total. These instances described scenarios with the potential for personnel exposure. 

The first instance described a measurement of an in-core filter that had been removed. It had a 

reading of 20 mrad/hr beta-gamma including 5 mr/hr gamma (Atomics International, 1964a, p. 17). 

The second instance described a survey that was taken inside the purification column and overheads 

cooler that had been opened for maintenance. The highest reading found at this location was 10 

mrads/hr beta-gamma, including 0.5 mrad/hr gamma at 1-inch” from the contaminated surface 

(Atomics International, 1964c, p. 19). 

Neutron

There is no complete set of data that explicitly covers neutron dose. However, based on the available 

summary data, the individual data reports (reported to the AEC as whole-body radiation exposures to 

penetrating radiation) (AEC, 1963; AEC, 1964; AEC, 1965a; AEC, 1966a) would likely have 

included neutron dose, had any been detected. In the summary data that were reported to the AEC 

Idaho Operations Office for 1963, the 31 Atomic International employees’ dosimeters indicated that 

they received 0 neutron dose that year (City of Piqua, 1964). 

NIOSH did not find any ambient neutron flux measurements for occupied areas. However, NIOSH 

did find estimates of the neutron flux based on activation products in the reactor side of the biological 

shield that were made prior to retirement of the facility (Hewson, 1969). The flux estimates, when 

extrapolated to the outer surface of the biological shield and multiplied by a generic conversion factor 

developed by the Nuclear Regulatory Commission for a neutron energy spectrum typical for reactors 

(Schwartz, 1983), indicate that the average dose at the outer surface of the shield was in the singledigit microrem per hour (μrem/hr) range. 
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During an interview, a former POMR health physics technician stated that neutron surveys were 

routine and performed initially each time the power level was raised (Personal Communication, 

2009g). One set of neutron results collected in 1964 was reported in a semiannual report (Atomics 

International, 1964a p. 83). Three locations were listed as having no neutrons, while two others were 

listed as having <0.5 mrem/hr neutrons and had a corresponding gamma dose rate. The survey was 

taken while the plant was operating at full power (i.e., 45.5 MWt). 

6.2.2 Available POMR External Monitoring Data for the Post-Operational Period 

Photon

Table 6-3 summarizes the dosimeter results for the post-operational period that were sent annually to 

the AEC Chicago Operations Office. 

Table 6-3: Summary of Annual Whole-Body Penetrating Exposures for the POMR Post-Operational Period 

Year No. Identified as 

NOT Monitored 

No. Identified as 

Monitored 

No. Identified with 

0-1 rem 

No. Identified with 

1-2 rem 

1966 0 50 42 8 

1967 0 48 47 1 

1968 0 36 36 0 

19691 - - - - 

Notes: 

1

 NIOSH has not found any data for 1969 that were sent to AEC Chicago Operations. 

Information for 1966 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1966 (AEC, 1966a). 

Information for 1967 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1967 (AEC, 1967). 

Information for 1968 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1968 (AEC, 1968). 

For 1966, the Piqua Nuclear Power Facility Reactor Operations Analysis Program Semiannual 

Progress Report No. 9 (July 1, 1966 - December 31, 1966) did not report the maximum external 

exposure at the site (Atomics International, 1966e). However, it did mention an event where the 

highly irradiated upper-core grid plate was raised above the coolant levels, creating high ambient 

radiation levels. The report indicated that "radiation readings did not exceed 5 r/hr at the reactor floor 

level" (Atomics International, 1966e, p. 10). After the reactor was drained of coolant and certain 

components were removed, it was visually inspected. The report stated that, "In general, a radiation 

field of 50 mr/hr existed at the periscope operator's position" (Atomics International, 1966e). 

A separate report, apparently describing the same events, indicated that the radiation level over the 

open reactor at floor level was 6 R/hr (Seyfrit, unknown date, p. 12). This report stated that during the 

transfer of the upper-core grid, lower-guide grid, and barrel from the reactor to a storage pool, 

radiation levels ranged from 500 mr/hr at 20 feet and 2 R/hr at about 2 feet. The report said exposure 

was "generally limited to the lesser rate" and the highest integrated dose was about 120 mr (Seyfrit, 

unknown date). 

For the year 1966, the site summary reports identified in Table 6-3 indicated that annual doses 

exceeded the 1 rem value, but were less than the 2 rem value for eight workers. 
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For the years 1967 and 1968, there was only one semiannual report available, Report Number 10, 

which covered the first half of 1967 (Atomics International, 1967b); it did not mention any personnel 

exposures. There was also only one monthly report available, Number 53, for September 1967 

(Atomics International, 1967a). It stated that during the month of August, 39 films were processed 

with 13 having exposures ranging "from 20 mrem to a maximum of 470 mrem with an average of 128 

mrem" (Atomics International, 1967a, p. 14). This is consistent with the range of exposure reported to 

the AEC as listed in Table 6-3. The September 1967 report also stated that 12 self-reading dosimeters 

were used during the month of September and these showed readings "ranging from 2 mrem to 225 

mrem" (Atomics International, 1967a). 

For the year 1969, no exposure data reported to the AEC from the Piqua Nuclear Power Facility have 

been located. In the Piqua Nuclear Power Facility Final Monthly Progress Report No. 69, it states 

that the film badge service was discontinued on January 13th and that all personnel monitoring films 

showed minimal exposure (Atomics International, 1969, p. 11). The Piqua Nuclear Power Facility 

Final Survey Report documented ambient radiation levels as of January 13, 1969 (Geiger, 1969). 

These levels can be taken to represent conditions for all of 1969, as the disassembly and 

decontamination had been completed in mid to late 1968 (Wheelock, 1970). Most areas were 

measured as 0.02 mrad/hr, with the highest listed reading as 0.8 mrad/hr at the sump wall (Geiger, 

1969, p. 3). The report also indicated that the average reading on this wall was 0.1 mrad/hr. Only two 

other locations were listed in the report as greater than 0.02 mrad/hr; these included the drain F4 at the 

56 foot level at 0.4 mrad/hr (Geiger, 1969, p. 7) and the fuel cask cable at the 100 foot level at 0.20 

mrad/hr (Geiger, 1969, p. 6). 

In the Chicago Operations Office Review of the Phase IIA Recovery Program Plan of Action, there 

was a requirement by the AEC for wrist badges to be worn during the recovery phase "any time 

personnel are in the vicinity of the vessel and there is the potential for radiation streaming" (AEC, 

1966b). In a draft plan for Phase IIa of the recovery phase, it stated that wrist badges would be worn 

by personnel working in areas of potentially high radiation-streaming fields (Morgan, 1966). No 

results for wrist badges in primary or summarized form were located by NIOSH. 

Table 6-4: Semiannual and Monthly Report Results for the POMR Post-Operational Period 

Table 6-4 spans two pages. 

Timeframe Results 

Report Type 

and Number Begin End 

Number of 

Positive 

Results 

Maximum 

Result 

Notes from the Radiation Monitoring 

Section of the reports 

Semiannual-8 

(Atomics 

International, 

1966b) 

01/01/66 06/30/66 Unknown Unknown 5 mR/h beta and gamma; sample of coke 

surface (Atomics International, 1966b, p. 

103). 

Semiannual-9 

(Atomics 

International, 

1966e) 

07/01/66 12/31/66 Unknown Unknown No section on plant radiation levels. 

Description of unplanned event where upper 

core-grid plate was raised above coolant 

levels creating radiation levels that "did not 

exceed 5 r/hr at the reactor floor level." 

Semiannual-10 

(Atomics 

International, 

1967b) 

01/01/67 06/30/67 Unknown Unknown No section on plant radiation levels. 

Monthly-53 09/01/67 09/30/67 39 (13 were 470 mrem N/A 
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Table 6-4: Semiannual and Monthly Report Results for the POMR Post-Operational Period 

Table 6-4 spans two pages. 

Timeframe Results 

Report Type 

and Number Begin End 

Number of 

Positive 

Results 

Maximum 

Result 

Notes from the Radiation Monitoring 

Section of the reports 

(Atomics 

International, 

1967a) 

above 20 

mrem, the 

remainder 

"showed 

minimal 

exposures") 

Final Monthly69 

(Atomics 

International, 

1969) 

02/01/69 02/28/69 Unknown Unknown N/A 

Beta

Although it is assumed that film badges used in the post-operational period would have had a beta 

component, no documents have been reviewed that indicate this specifically or contain any results of 

the monitoring. 

Neutron

No neutron monitoring data were discovered for the non-operational period at the site, as the reactor 

was shut down during the non-operational period and many activities were performed with the fuel 

removed from the reactor core. Based on this information, NIOSH does not expect that there would 

have been any appreciable neutron exposures associated with the post-operational period activities. 

Therefore, further evaluation of potential neutron exposures are not included for the post-operational 

period evaluated in this report. 

7.0 Feasibility of Dose Reconstruction for the Class Evaluated by 

NIOSH 

The feasibility determinations for the class of employees under evaluation in this report are governed 

by both EEOICPA and 42 C.F.R. § 83.13(c)(1). Under that Act and rule, NIOSH must establish 

whether or not it has access to sufficient information either to estimate the maximum radiation dose 

for every type of cancer for which radiation doses are reconstructed that could have been incurred 

under plausible circumstances by any member of the class, or to estimate the radiation doses to 

members of the class more precisely than a maximum dose estimate. If NIOSH has access to 

sufficient information for either case, NIOSH would then determine that it would be feasible to 

conduct dose reconstructions. 

In determining feasibility, NIOSH begins by evaluating whether current or completed NIOSH dose 

reconstructions demonstrate the feasibility of estimating with sufficient accuracy the potential 
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radiation exposures of the class. If the conclusion is one of infeasibility, NIOSH systematically 

evaluates the sufficiency of different types of monitoring data, process and source or source term data, 

which together or individually might assure that NIOSH can estimate either the maximum doses that 

members of the class might have incurred, or more precise quantities that reflect the variability of 

exposures experienced by groups or individual members of the class. This approach is discussed in 

OCAS’s SEC Petition Evaluation Internal Procedures which are available at 

http://www.cdc.gov/niosh/ocas. The next four major subsections of this Evaluation Report examine: 

 The sufficiency and reliability of the available data. (Section 7.1) 

 The feasibility of reconstructing internal radiation doses. (Section 7.2) 

 The feasibility of reconstructing external radiation doses. (Section 7.3) 

 The bases for petition SEC-00126 as submitted by the petitioner. (Section 7.4) 

7.1 Pedigree of POMR Data 

This subsection answers questions that need to be asked before performing a feasibility evaluation. 

Data Pedigree addresses the background, history, and origin of the data. It requires looking at site 

methodologies that may have changed over time; primary versus secondary data sources and whether 

they match; and whether data are internally consistent. All these issues form the bedrock of the 

researcher’s confidence and later conclusions about the data’s quality, credibility, reliability, 

representativeness, and sufficiency for determining the feasibility of dose reconstruction. The 

feasibility evaluation presupposes that data pedigree issues have been settled. 

Available data for the POMR facility are in the form of summary reports, including monthly, 

semiannual, and annual reports for the years 1963 through 1969. The types of monitoring data 

reported in the monthly and semiannual reports are inconsistent from report to report. If NIOSH were 

to only use these reports, there would be some information gaps for all years of operation. NIOSH 

does not have access to the individual, hardcopy monitoring data. 

7.1.1 Internal Monitoring Data Pedigree Review 

The following subsections summarize the internal monitoring data pedigree for the operational and 

post-operational periods at the POMR facility. 

7.1.1.1 Internal Monitoring Data Pedigree Review for the Operational Period

ATTRIBUTION: Section 7.1.1.1 was completed by Daniel Mantooth, Dade Moeller and Associates, 

Inc. These conclusions were peer-reviewed by the individuals listed on the cover page. The 

rationales for all conclusions in this document are explained in the associated text. 

No original records for bioassay, area air samples, or surface contamination have been located for the 

operational period. There are several cases where air sampling results (reported as maximum, 

minimum, and average) for stack effluent and environmental activity levels are reported in summary 

documents. Other documents report these parameters and workplace air levels subjectively (i.e., < 

MPC). A data quality review cannot be performed on the summary data that was reported to the AEC 

without the actual data records. The summary documents reviewed consist of various monthly, 

quarterly, and semiannual reports of reactor plant conditions that collectively cover the entire thirtytwo month period in which the reactor operated. Although this information is not the original data, 
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the information contained within these documents, when taken together, indicates that POMR 

radiological conditions were benign with respect to internal dose; an indication supported by the 

reports of former POMR personnel interviewed during the course of this evaluation report (Personal 

Communication, 2009a; Personal Communication, 2009e; Personal Communication, 2009g). 

Collectively, every source reviewed and interview conducted leads to the conclusion that surface and 

airborne contamination levels and radionuclides present during POMR operations were unlikely to 

have resulted in measurable internal dose. 

7.1.1.2 Internal Monitoring Data Pedigree Review for the Post-Operational Period

ATTRIBUTION: Section 7.1.1.2 was completed by Eugene W. Potter, M. H. Chew & Associates, Inc. 

These conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for 

all conclusions in this document are explained in the associated text. 

No original records for bioassay, area air samples, or surface contamination have been located for the 

post-operational period. There are reports similar to those described in Section 7.1.1.1; however, only 

a single monthly report with stack, environmental or on-site air concentrations was located (Atomics 

International, 1967a). Although some information was identified as being associated with the preand post-decontamination and decommissioning activities, the information is not sufficient to support 

the evaluation of internal exposures for the post-operational period (Wheelock, 1970; Atomics 

International, 1967a). 

7.1.2 External Monitoring Data Pedigree Review 

The following subsections summarize the external monitoring data pedigree for the operational and 

post-operational periods at the POMR facility. 

7.1.2.1 External Monitoring Data Pedigree Review for the Operational Period

ATTRIBUTION: Section 7.1.2.1 was completed by Louise Buker, Oak Ridge Associated Universities 

(ORAU). These conclusions were peer-reviewed by the individuals listed on the cover page. The 

rationales for all conclusions in this document are explained in the associated text. 

Although external monitoring did occur at POMR, NIOSH has not yet been able to locate the original 

data records. The only external monitoring data available to NIOSH consist of summary dosimetry 

data reported annually to the AEC offices of Chicago Operations and Idaho Operations. Without the 

actual data records, a data quality review cannot be performed on the summary data that were reported 

to the AEC. 

However, there is corroborating evidence that supports NIOSH’s ability to bound the external dose for 

the operational period evaluated in this report. A 1963 summary report to the AEC Idaho Operations 

Office indicates that 30 Atomics International workers who worked at the POMR facility received 0 

gamma, 0 beta, and 0 neutron dose for the year 1963, and one Atomics International worker received 

60 mrem gamma, 20 mrem beta and 0 mrem neutron (City of Piqua, 1964). In addition, former 

POMR personnel interviewed during the course of this evaluation report provided information about 

the POMR facility’s radiation protection practices that coincides with the available summary data. 

Information obtained from interviews with former POMR facility personnel indicates that (1) 

personnel working at the site were issued monthly or quarterly film badges, and (2) area surveys were 
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conducted on a routine and/or as-needed basis. Also, the monitoring was conducted for beta, gamma, 

and neutron radiation. NIOSH has compared the actual result information (as relayed in Section 6.2.1, 

and contained in Table 6-2 of this report) to the maximum values in the ranges of doses listed in the 

summary data reports for the respective operational years. This comparison supports NIOSH’s 

conclusion that the summary report data provided in Table 6-1 can be used as a bounding estimate for 

the external doses for the operational period. 

7.1.2.2 External Monitoring Data Pedigree Review for the Post-Operational Period

ATTRIBUTION: Section 7.1.2.2 was completed by Roger Halsey, Oak Ridge Associated Universities 

(ORAU). These conclusions were peer-reviewed by the individuals listed on the cover page. The 

rationales for all conclusions in this document are explained in the associated text. 

As reported for the operational period, although external monitoring did occur at POMR, NIOSH has 

not yet been able to locate the original data records. The only external monitoring data available to 

NIOSH consist of summary dosimetry data reported annually to the AEC offices of Chicago 

Operations and Idaho Operations. Without the actual data records, a data quality review cannot be 

performed on the summary data that were reported to the AEC. 

Unique to the POMR post-operational period was the potential for exposure to uneven or streaming 

gamma fields. This exposure risk was recognized and planned for through AEC requirements (AEC, 

1966b). In addition, the use of wrist badges during recovery-program operations with the potential for 

"radiation streaming" was written into planning documents (Morgan, 1966). NIOSH has not been 

able to locate any wrist badge results or confirm that wrist badges were actually worn. The only 

reference to any extremity monitoring in the employee interviews was a determination (at the 

beginning of operations) that ring badges were initially used when "handling" the organic moderator 

and were found to be unnecessary (Personal Communication, 2009g). 

In light of the evaluation conclusions reached in this report regarding internal dose reconstruction 

feasibility for the post-operational period at POMR, NIOSH did not perform an extensive external 

data sufficiency and pedigree evaluation for external data for the post-operational period. NIOSH 

intends to use the available external monitoring data, and dose reconstruction approaches defined for 

the operational period to support partial dose reconstructions (to include medical X-ray exposures) for 

the post-operational period. 

7.2 Evaluation of Bounding Internal Radiation Doses at POMR 

ATTRIBUTION: Section 7.2 and its related subsections were completed by Daniel Mantooth, Dade 

Moeller and Associates, Inc. and Eugene W. Potter, M. H. Chew & Associates, Inc. These 

conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for all 

conclusions in this document are explained in the associated text.

The principal source of internal radiation doses for members of the class under evaluation was 

exposure to beta/gamma emitting radionuclides from four sources: (1) activated impurities in the 

coolant, (2) activated corrosion products, (3) neutron recoil reactions with the aluminum cladding, and 

(4) tritium produced by ternary fission. The following subsections address the ability to bound 

internal doses, methods for bounding doses, and the feasibility of internal dose reconstruction. 
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7.2.1 Evaluation of Bounding Internal Doses during the Operational Period 

7.2.1.1 Evaluation of Bounding Process-Related Internal Doses during the Operational Period 

NIOSH was unable to locate any urinalysis data, whole-body count data, or any other type of bioassay 

data for the operational period. Detailed below is a summary of the particulate airborne activity 

information available for reconstructing the process-related internal doses of members of the class 

under evaluation during the operational period. 

As discussed previously, particulate radionuclides consisting of sodium-24, phosphorus-32, sulfur-35, 

chlorine-38, chromium-51, manganese-54, iron-55, manganese-56, iron-59, arsenic-76, magnesium27, cobalt-58, and cobalt-60 were detected in the organic coolant, in in-vessel filters, and in filters 

associated with the coolant purifications system, the degasifications system, the fuel storage poolpurification system, miscellaneous residues and fluids, carbonaceous deposits, and the waste-fired 

boiler. It was proposed that tritium and carbon-14 would be present in the coolant at levels up to 0.21 

μCi/cc and 1.4E-04 μCi/cc, respectively (Atomics International, 1961). Airborne activity, if present, 

would have been generated by maintenance activities on these systems. As mentioned previously, no 

work area air sampling data have been located that could be used to support establishing bounding 

estimates of the airborne activity. 

The Final Safeguards Summary Report for the Piqua Nuclear Power Facility (Atomics International, 

1961) provides calculations demonstrating that operational airborne levels would be less than the 

MPC specified in the 1960 regulations that were current at the time (10 C.F.R. pt. 20). Summary 

documents reported air sampling results in relative terms (i.e., “...no personal contaminations or 

inhalations” and “...airborne levels in containment building were less than background”) that 

corroborate the Final Safeguards Summary Report for the Piqua Nuclear Power Facility (Atomics 

International, 1964a; Atomics International, 1965a; Atomics International, unknown date-a; Atomics 

International, 1965b). Because the reactor was a closed-loop primary system, and the coolant became 

a solid at room temperature, it is unlikely that personnel contamination and inhalation were factors. 

Air monitoring capability is described in Compilation of Piqua Nuclear Power Facility Operating 

Limits and Controls and Post-Critical Operational Tests, and confirms that the air sampling 

equipment (with the exception of tritium and carbon-14) could detect total beta-gamma activity at the 

MPC level for cobalt-60 (9E-9μCi/cm3) for restricted areas and would alarm before activity 

approached hazardous levels (Atomics International, 1965a). The reported sensitivity for workplace 

air monitors for particulate activity was 1E-12 μCi/cm3 (Atomics International, 1965a), or nearly 3 

orders of magnitude less than the MPC. The presence of workplace air samplers, their capabilities, 

and consistently low airborne activity levels were also confirmed by former Piqua personnel (Personal 

Communication, 2009a; Personal Communication, 2009e; Personal Communication, 2009g). 

There were no incidents reported that would have led to significant personnel contamination. As an 

example of the type of incidents reported, Piqua Nuclear Power Facility Reactor Operations Analysis 

Program Semiannual Progress Report No. 6 (January 1-June 30, 1965) mentioned that the soot 

collection bag from the waste-fired boiler came loose and released contaminated soot into the 

auxiliary building. Total surface contamination levels of 1,000 dpm/44 in2

 (~400 dpm/100 cm2) were 

measured (Atomics International 1965c, p. 77). 
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No personnel contaminations or inhalations occurred when the soot collection bag came loose. The 

radioisotopes involved were not specified, but it can be assumed that they were comprised of those 

described throughout this report and listed in Table 5-3. Based on its research to date, NIOSH has not 

located air sampling data sufficient by themselves to support bounding the process-related internal 

dose. However, the information discussed previously and further evaluated throughout the rest of this 

report, supports the conclusion that airborne particulate activity in process areas during the operational 

period was not likely to exceed the MPCs. The bounding process-related internal doses from the 

particulate radionuclides can be obtained using the MPC values. Intakes of tritium and carbon-14 are 

calculated using the maximum ratio of the reported or measured activities in the coolant or process 

residues, respectively (see Section 5.2.2) to the total reported particulate activity. The resulting ratio 

is then applied to the expected process airborne level (i.e., MPC) to derive the bounding value for 

tritium and carbon-14. 

7.2.1.2 Evaluation of Bounding Ambient Environmental Internal Doses during the Operational 

Period 

Ambient environmental internal doses could have resulted from the inhalation of radionuclides that 

were exhausted from the reactor plant. Ambient air particulate levels were monitored in stack 

effluents at both on- and off-site (so-called “environmental samples”) locations. Data for stack 

particulates have not been located, but several reports indicate that stack effluents were consistently 

“...less than the MPC” (Atomics International, unknown date-a; Atomics International, 1964a; 

Atomics International, 1964b; Atomics International, 1964c; Atomics International, 1965c). In 

addition, the stack monitoring system was reported to have sufficient sensitivity to detect particulate 

activity equal to 3E-10 μCi/cm3

 of iodine-131 (Atomics International, 1965a), which is equivalent to 

the unrestricted MPC for cobalt-60, the limiting radionuclide of concern at the POMR facility. Since 

the maximum beta energies for iodine-131 and cobalt-60 are on the same order (i.e., 0.606 and 0.318), 

it can be assumed that the monitoring instrument sensitivity would be similar. Alarms would sound in 

the control room when activity exceeded the MPC level. The information regarding stack effluent 

values and monitoring coincides with the estimated values provided in the Final Safeguards Summary 

Report for the Piqua Nuclear Power Facility (Atomics International, 1961). In the case of onsite and 

offsite monitoring, several monthly reports provide results of this sampling, all indicating a total 

particulate activity well below the outside operational (unrestricted area) MPC values (Atomics 

International, 1964c; Atomics International, 1965b; Atomics International, 1965d). 

Based on its research to date, NIOSH has located limited air sampling data to support bounding 

ambient environmental internal dose. However, the information discussed previously supports 

NIOSH’s conclusion that airborne activity levels outside the process areas and the plant grounds can 

be bounded by application of operational internal dose assessment methods. 

7.2.1.3 Methods for Bounding Process-Related Internal Doses during the Operational Period

The method for bounding the operational period internal dose is based on the conclusion that the 

airborne activity in operational areas would not have exceeded the lowest applicable MPC without 

being detected by the existing monitoring system and noted in the routine summary reports. This 

conclusion is based on information in the SRDB and interviews with former POMR facility personnel 

indicating that the POMR facility had, and used, workplace alarming airborne monitoring equipment 

capable of detecting airborne particulate activity to the MPC levels in effect at the time (as noted in 10 

C.F.R. pt. 20, Appendix B Table II). Typically, the alarm on air monitors is set at some percentage of 
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the applicable limit for the most restrictive radionuclide (i.e., lowest MPC). For the purposes of 

bounding the operational period internal dose, it will be assumed that the alarms were set to trip at the 

MPC for the most restrictive radionuclide. The radionuclide with the lowest MPC listed in Table 5-3 

is chlorine-38 (3E-09 μCi/cm3

). However, due to its short half-life (37.3 min), it is likely that 

chlorine-38 levels could have largely decayed before sufficient activity would have been collected by 

the air monitor to cause the alarm to signal. The radionuclide with the next lowest MPC is cobalt-60 

(9E-09 μCi/cm3

); with a half-life of over 5 years, sufficient activity would collect to reach alarm 

levels if elevated airborne activity was present. If the total beta/gamma airborne activity did not 

exceed the MPC for cobalt-60 (9E-09 μCi/cm3

), and the total particulate activity consisted of 48% 

impurities, 13% corrosion products, and 39% recoil products, the total air activity would be comprised 

of 4E-09 μCi/cc impurities, 1E-09 μCi/cc corrosion products, and 4E-09 μCi/cc recoil products. 

These concentrations can be used to assess the radionuclide exposures in each applicable category (see 

Table 7-1). The radionuclide that represents the maximum internal dose contributor in each category 

can be evaluated on a case-by-case basis, and the categories can be combined to establish a composite 

dose estimate. The application of this method supports NIOSH’s ability to bound the internal dose 

from exposure to these radionuclides for the evaluated class during the operational period. 

To calculate the airborne levels of tritium and carbon-14 in process areas, ratios were calculated 

between the maximum expected or reported activity of these two radionuclides to the total activity in 

the coolant and residues from other radionuclides (activation, corrosion, and recoil), as described 

previously. Information obtained from several documents provide data indicating an average coolant 

activity over a wide range of conditions of 1.1E-02 μCi/cm3

 (Atomics International, 1964b; Atomics 

International, 1964a; Atomics International, 1965d; Atomics International, 1965b). The greatest 

tritium activity discovered in the available documentation is the estimate provided in the Final 

Safeguards Summary Report for the Piqua Nuclear Power Facility, of 0.21 μCi/cm3. The ratio of this 

tritium activity to the total activity would be 19.1 (0.21 μCi/cm3 ÷ 1.1E-02 μCi/cm3 = 19.1). Since the 

maximum airborne activity is not expected to exceed the MPC for cobalt-60 (9E-09 μCi/cm3

), the 

tritium airborne concentration should not exceed 1.7E-07 μCi/cm3 (19.1 x 9E-09 μCi/cm3). The 

maximum activity value found for carbon-14 was reported in the analyses of soot from the flue gas 

collector , which is equal to 2.1E-02 μCi/cm3 (Wheelock, 1970 Appendix B, Table B-2). The ratio of 

this activity to the total activity is 1.9(2.1E-02 μCi/cm3 ÷ 1.1E-02 μCi/cm3 = 1.9), which would result 

in a process airborne concentration of 1.7E-08 μCi/cm3

 (1.9 X 9E-09 μCi/cm3 = 1.7E-08 μCi/cm3). 

These concentrations can be used to assess the respective radionuclide exposures. The application of 

this method supports NIOSH’s ability to bound the internal dose from exposure to these radionuclides 

for the evaluated class during the operational period. 

Table 7-1: Radionuclides of Concern by Category 

Coolant Impurities/ 

Components 

Corrosion 

Products Recoil Products 

Na-24 Cr-51 Na-24 

P-32 Mn-54 Co-58 

S-35 Fe-55 Co-60 

Cl-38 Mn-56 - 

Mn-56 Fe-59 - 

As-76 - - 

Tritium - - 

C-14 - - 
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The assessment methods presented here define the methods by which a maximum internal dose 

estimate can be determined for the evaluated class, which supports NIOSH’s conclusion that the 

operationally-related internal dose during the operational period can be bound. 

7.2.1.4 Methods for Bounding Ambient Internal Doses during the Operational Period

The method for bounding the operational period ambient environmental internal dose is based on the 

conclusion that the total airborne activity in non-process areas would not have exceeded the lowest 

applicable non-process MPC without being detected by the existing monitoring system and noted in 

the routine summary reports. This conclusion is based on information in the SRDB and interviews 

with former POMR facility personnel, which indicated that the POMR facility had, and used, 

environmental and onsite sampling and analysis procedures with sufficient sensitivity to detect 

airborne particulate activity to the MPC levels in effect at the time (as noted in 10 C.F.R. pt. 20, 

Appendix B Table II). Typically, action levels for environmental monitoring are established at some 

percentage of the applicable limit for the most restrictive radionuclide (i.e., lowest MPC). For the 

purpose of bounding the ambient environmental dose, it will be assumed that the action level is equal 

to the MPC for the most restrictive radionuclide; the radionuclide in Table 5-3 with the lowest MPC is 

cobalt-60 (3E-10 μCi/cm3

 [10 CFR pt 20, 1961]). If the total beta/gamma airborne activity did not 

exceed 3E-10 μCi/cm3

, and the total particulate activity consisted of 48% impurities, 13% corrosion 

products, and 39% recoil products, the total air activity would be comprised of 1E-10 μCi/cc 

impurities, 4E-11 μCi/cc corrosion products, and 1E-10 μCi/cc recoil products. 

The ratios calculated in Section 7.2.1.3 (between the expected activity of tritium and carbon-14 and 

the total measured coolant activity) are used to determine ambient environmental airborne levels. For 

tritium, the expected airborne activity would be 5.7E-09 μCi/cm3

 (19.1 x 3E-10 μCi/cm3). For 

carbon-14, the expected ambient airborne level would be 5.7E-10 μCi/cm3 (1.9 x 3E-10 μCi/cm3 = 

5.7E-10 μCi/cm3

). 

As discussed in Section 7.2.1.2 and presented in this assessment of the ambient environmental 

airborne concentrations (as compared to the operationally-related values), the ambient environmental 

dose, based on the airborne activity levels outside the process areas and the plant grounds (and 

associated internal doses), can be bounded by the operationally-related values/levels and internal dose 

assessment methods. 

7.2.1.5 Internal Dose Reconstruction Feasibility Conclusion for the Operational Period

An extensive review of the available data, as well as interviews with former POMR facility personnel, 

have revealed no evidence of a radiological condition, transient or otherwise, that would lead to 

significant internal dose for the class evaluated in this report during the operational period. In 

addition, information was located that leads to the conclusion that workplace monitoring, with the 

exception of tritium and carbon-14, was sufficiently sensitive to detect airborne radioactivity at levels 

lower than the MPC values in effect during the operational period for the facility. The tritium and 

carbon-14 concentrations were evaluated through the determination of coolant concentration ratios, 

based on what would be considered worst-case calculated conditions as defined in the Final 

Safeguards Summary Report for the Piqua Nuclear Power Facility (Atomics International, 1961). 

Based on this information and the assessment as presented in Section 7.2.1 of this report, NIOSH has 

concluded that it is feasible to bound the internal dose (reconstruct dose with sufficient accuracy) for 

the class evaluated in this report during the operational period. 
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7.2.2 Evaluation of Bounding Internal Doses during the Post-Operational Period 

7.2.2.1 Evaluation of Bounding Process-Related Internal Doses during the Post-Operational Period 

NIOSH was unable to locate any urinalysis data, whole-body count data, or any other type of bioassay 

data for the post-operational period. The radionuclides of concern are the same as those during the 

operational period. 

During the post-operational period, some work activities were similar to those in the operational 

period (e.g., changing filters, de-fueling, burning coolant, etc.). However, there were other nonroutine activities that were more likely to generate airborne radioactivity during the post-operational 

period, such as drilling, welding, cutting, working inside the reactor performing modifications on 

reactor-core components, sampling/handling coke, etc. Additional examples of recovery period 

activities include the following: 

 Removing stuck rods/fuel assemblies; 

 Removing fuel fouled with a film layer; 

 Profiling the core (mapping coke mass); 

 Draining all coolant; 

 Removing carbonaceous material; 

 Sampling and handling samples of coke in a “vented glovebox”; 

 Inspecting the reactor core and fuel assemblies; 

 Removing the upper core grid assembly and inner barrel assembly, using the polar crane; 

 Modifying in-core components; 

 Drilling (air powered) three 3⁄4-inch holes in the lower-grid supporting barrel; and 

 Working on an in-core work platform. 

The primary activities went from closed-loop operations to open vessel/core activities in order to cool 

the system down and perform in-core modifications. The reactor vessel head was removed, and the 

core was opened to the containment for an extended period of time. In addition, some of the core 

components (e.g., upper-core grid plate and support barrel) were removed for the purposes of viewing 

the inside of the reactor vessel, inspecting/removing the carbonaceous mass, and performing core 

modifications to improve flow properties in the core. There was an extensive effort to clean up the 

coke mass in the core. This work was mostly performed over the top of the reactor with pool-type 

manipulators. Some interviews discussed looking directly into the reactor (Personal Communication, 

2009i; Personal Communication, 2009j). 

During the decontamination and decommissioning period, piping was cut by flame and mechanical 

methods. The reactor outlet and inlet piping had to be cut and seal welded afterwards. Although the 

piping and auxiliary systems were flushed, there was reference to activity for cobalt-60, tritium, and 

carbon-14 still being a significant internal radiation hazard. While coolant/moderator was burned in 

the waste-fired boiler as a part of normal operations, during the post-operational period a substantial 

amount of coolant was disposed of in this manner. 

The procedures used during the operational period had to be modified to conduct this work. For 

example, some fuel handling and fuel inspection activities (normally performed solely with the fuel-
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handling machine) were conducted with the overhead crane. Many of the post-operational activities

were non-routine, and there was certainly an increase for internal hazards for POMR personnel during 

the post-operational period. The post-operational activities were more hands-on than those in the 

operational period. During the operational period, health physicists could anticipate the airborne 

concentrations from experience. This would not have been the case for the post-operational period. 

References were found indicating the need for local ventilation (Ashley, 1966; AEC, 1966c) and 

supplied-air (Atomics International, unknown date-b, pp. 3-8; Ashley, 1966). The plans stated that air 

sampling would be performed as required, that appropriate respiratory protection devices would be 

utilized as required, and that bioassay samples would be collected if required by air sampling data or 

personnel contamination results (Atomics International, 1968c, p. 11). However, NIOSH has not 

found results of work zone air samples or any verification that such samples were taken. Based on 

this assessment, NIOSH has concluded that, given the increased potential for internal radiological 

exposures associated with the post-operational period, NIOSH does not have sufficient information or 

data to support bounding the internal dose related to recovery and decontamination and 

decommissioning activities during the post-operational period.

7.2.2.2 Evaluation of Bounding Ambient Environmental Internal Doses during the Post-Operational 

Period 

Ambient environmental internal doses could have resulted from the inhalation of radionuclides 

exhausted from the reactor plant since ventilations systems remained intact during the post-operational 

period. Ambient air particulate levels were monitored in stack effluents at both on- and off-site (socalled “environmental samples”) locations. Data for stack particulates have not been located, and only 

a single report indicates that stack effluents were at background levels (Atomics International, 1967a). 

The stack monitoring system in place during the post-operational period had the same sensitivity to 

detect particulate activity as during the operational period. 

NIOSH has only located air sampling data for a one month to support bounding ambient 

environmental internal dose during the post-operational period. Therefore, NIOSH has concluded that 

there is not sufficient information to conclude that airborne activity levels outside the process areas 

and the plant grounds can be bounded, either by evaluation of the available data or by application of 

process related internal dose assessment methods, for the post-operational period evaluated in this 

report. 

7.2.2.3 Methods for Bounding Process-Related Internal Doses during the Post-Operational Period

Following the same logic as for the operational period, the post-operational period operations related 

internal dose could be based on the conclusion that the airborne activity in operational areas would not 

have exceeded the lowest applicable MPC without being detected by the existing monitoring system 

and noted in the routine summary reports. However, for the post-operational period only a single 

monthly report (September 1967) with stack, environmental, and on-site filter results was located 

(Atomics International, 1967a). The activities conducted were non-standard and were likely to have 

resulted in increased airborne concentrations. Plans for the post-operational period activities 

recognized the need for local ventilation and respiratory protection, but there were no reports located 

that detailed the actual use of these measures. NIOSH concludes that the information recovered is 

insufficient to bound the internal doses during the post-operational period. 
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7.2.2.4 Methods for Bounding Ambient Internal Doses during the Post-Operational Period

Following the same logic as for the operational period, the post-operational period ambient 

environmental internal dose could be based on the conclusion that the airborne activity in non-process 

areas would not have exceeded the lowest applicable non-process MPC without being detected by the 

existing monitoring system and noted in the routine summary reports. However, only a single 

monthly report with stack, environmental, and on-site filter results for the post-operational period was 

located (Atomics International, 1967a). NIOSH concludes that this information is insufficient to 

bound the ambient environmental internal doses during the post-operational period. Although no 

specific ambient environmental data exist to support the bounding assessment of this dose for the 

post-operational period in this evaluation, NIOSH will apply any available personnel monitoring data 

that may exist (on a case-by-case basis) in support of a partial dose reconstruction for claims with 

non-presumptive cancers and those with less than 250 days of employment. 

7.2.2.5 Internal Dose Reconstruction Feasibility Conclusion for the Post-Operational Period

NIOSH has learned that there was a greater potential for airborne radioactivity, but less air 

concentration information has been located for the post-operational period. Based on this information 

and the assessment as presented in Section 7.2.2 of this report, NIOSH has concluded that it is not 

feasible to bound the internal dose (reconstruct dose with sufficient accuracy) for the class evaluated 

in this report for the post-operational period. 

7.3 Evaluation of Bounding External Radiation Doses at POMR 

ATTRIBUTION: Section 7.3 and its related subsections were completed by Louise Buker, Oak Ridge 

Associated Universities (ORAU) and Roger Halsey, Oak Ridge Associated Universities (ORAU). 

These conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for 

all conclusions in this document are explained in the associated text. 

The principal sources of external radiation doses for members of the proposed class were the Piqua 

Organic Moderated Reactor and the calibration sources used to calibrate the radiation monitoring 

equipment. 

The following subsections address the ability to bound external doses, methods for bounding doses, 

and the feasibility of external dose reconstruction. 

7.3.1 Evaluation of Bounding External Doses during the Operational Period 

The following subsections summarize the extent and limitations of information available for 

reconstructing the external doses during the operational period for members of the class under 

evaluation. 

7.3.1.1 Personnel Dosimetry Data for the Operational Period

The available external whole-body dosimetry data for the site is in the form of summary data reported 

to the AEC Chicago Operations office for the City of Piqua workers and to the AEC Idaho Operations 

for the Atomics International workers who worked at the POMR facility. This summary data cover 

the complete timeframe and are based on film badge data. For the purpose of bounding the possible 
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external doses, the external data appear to be reliable in summary form. The data are presented in 

Table 6-1 of this report. According to the references reviewed and an interview with a former Piqua 

health physics technician, the film badge exchange frequency was monthly for operation and 

maintenance personnel, and quarterly for administrative personnel (Personal Communication, 2009g). 

In addition to the summary data, there are three monthly reports that state the maximum exposure for 

the month, and a complete set of semiannual reports that make reference to employee external 

radiation exposures. Although the data are not fully comprehensive, whenever the maximum levels 

are mentioned, the maximum levels are much lower than the values in the AEC summary reports for 

the same time period. 

Photon

The data are sufficient to support bounding whole-body photon dose estimates for the years 1963 

through 1966 from the upper bound of 1 rem for the years 1963 through 1965, and 2 rem for the year 

1966; see Table 6-1 of this report. 

Beta

There is evidence that film badge beta monitoring was performed during the operational period, as 

indicated in the summary data. External dosimetry for plant workers were film badges provided by R. 

S. Landauer Jr. and Company (Atomics International, 1965b, p. 23), which included a shallow dose 

component. 

Documentation of shallow dosimetry results is incomplete for the operational lifetime of the POMR. 

In all of the documents that have been reviewed (City of Piqua, 1964), there is only one report of beta 

dose where gamma exposure is discussed (i.e., 60 mrem gamma and 20 mrem beta). 

Neutron

There is evidence that POMR facility personnel were monitored for neutron exposure with film 

badges (City of Piqua, 1964). During an interview, a former plant health physics technician said that 

he did not remember any positive neutron results from the Nuclear Track Emulsion (NTA), Type A 

film badges (Personal Communication, 2009g). Although there are no complete sets of neutron film 

badge results available, it is assumed that the complete set of summary data (reported to the AEC as 

whole-body radiation exposures from penetrating radiation) would have included neutron dose if it 

had been detected (AEC, 1963; AEC, 1964; AEC, 1965a; AEC, 1966a). 

The limitation of using this penetrating external exposure summary data as bounding is due to the lack 

of sensitivity of NTA film dosimeters at lower energies (Fix, 1997); NTA film has a lower energy 

threshold (~500 keV, below which the measured dose rapidly decreases). The threshold of ~500 keV 

could cause the dosimeter to miss much of the neutron dose in highly shielded workplace radiation 

fields, including those in a reactor facility. The shielding significantly reduces the neutron energy, 

altering the exposure spectrum. In general, if the calibration spectra is the same as the workplace 

spectra then an accurate neutron dose can be measured (i.e., the same fraction of the neutron spectra is 

missed in the calibration and in the workplace). However, in highly shielded facilities, the workplace 

neutron spectra are much more degraded than the calibration spectra (Fix, 1997, pp. 39-45). 
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Since the sensitivity of the NTA film is regarded as likely underestimating the neutron dose, the 

paired measurements of neutron and photon dose will be used to establish a neutron-to-photon ratio. 

Survey measurements reported in Piqua Nuclear Power Facility Operations Analysis Program 

Progress Report No. 4 (January 1, 1964-June 30, 1964) showed no detectable neutrons in normallyaccessible areas. In areas where neutrons were detected, the levels were low when compared to the 

potential for gamma exposure. Two measured values, each indicated as <0.5 mrem/hr, were found. 

These <0.5 mrem/hr readings were taken at the “OC-104, 14” line in the P-1B room location with a 

concurrent gamma reading of 13.5 mR/hr, and at the “OC-101, 14” line in the P-1A room location 

with a concurrent gamma reading 11 mR/hr (Atomics International, 1964a, p. 83). 

The survey (reported in Piqua Nuclear Power Facility Operations Analysis Program Progress Report 

No. 4 (January 1, 1964-June 30, 1964 ) was conducted when the plant was operating at full power. 

The first detectable reading was associated with a gamma level of 13.5 mR/hr, assuming a worst case 

neutron reading of 0.5 mrem/hr, with a resulting neutron-to-photon ratio of 3.7% (or 0.037 to 1). The 

second detectable reading was associated with a gamma level of 11 mR/hr, with a worst case neutron 

reading of 0.5 mrem/hr, with a resulting neutron-to-photon ratio of 4.5% (or 0.045 to 1). Based on 

this information, a neutron-to-photon ratio of 10% (or 0.10 to 1) can be applied as a bounding ratio (in 

round numbers) to assess neutron dose given a photon dose (which would be assigned based on the 

bounding photon dose approach already discussed). Since these measurements were made with a long 

counter (Personal Communication, 2009g), which is known for being neutron energy 

independent (Fix, 1997; Cember, 1969), this approach avoids any energy dependence issues related to 

the NTA badges. 

7.3.1.2 Area Monitoring Data for the Operational Period

Area monitoring data are available from progress reports, monthly reports, and semiannual reports. 

Photon

In the monthly, semiannual, and annual reports, there are references to ambient gamma levels, 

primarily presented as example levels. For example, Piqua Nuclear Power Facility Operations 

Analysis Program Progress Report No. 4 (January 1, 1964-June 30, 1964) indicated that “Gamma 

radiation which penetrated the biological shield was less than 1 mr/hr” (Atomics International, 1964a, 

p. 78). However, there are no consistent surveys or summary data available. 

The most complete set of monitoring data available is in a table summarizing a survey taken in 1964 

(during the first year of operation), described as a “typical survey.” The table includes readings from 

eighteen locations that ranged from 0.1 mr/hr at the “FELS Tank” location to 250 mr/hr at the “F-2A 

Degasifier filter” location (Atomics International, 1964a, p. 83). The report indicated that with the 

exception of the degasifier filter, radiation levels remained low enough to permit operator access for 

contact maintenance on the primary coolant system without time limits (Atomics International, 1964a, 

p. 82). 

Beta

There are several discrete cases mentioned in the reviewed documents where beta-gamma radiation 

levels were measured (Atomics International, 1967a, p. 15; Atomics International, 1965c, p. 76; 

Atomics International, 1964a, p. 17). NIOSH discovered only a couple of instances where the beta-
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gamma reading was distinguished from the gamma activity (Atomics International, 1964a, p. 17; 

Atomics International, 1964c, pp. 19-20). 

Neutron

In an interview with a former POMR plant health physicist technician, the former employee stated that 

neutron surveys were routinely performed and were performed at every power level, and that the type 

of detector used at the plant was a long counter (Personal Communication, 2009a; Personal 

Communication, 2009g) 

One of these surveys, conducted while the plant was operating at full power, was reported in Piqua 

Nuclear Power Facility Operations Analysis Program Progress Report No. 4 (January 1, 1964-June 

30, 1964); this report included a summary table listing the results of a “typical radiation survey” of the 

plant when it was operating at 45.4 MWt (Atomics International, 1964a). The neutron radiation 

surveys of the “Reactor 100-ft level” location, the “East Reactor Biological Shield at the 78-ft level” 

location, and the “East Reactor Shield at the 72-ft level” location were indicated as having “no 

neutrons” and were recorded as < 0.5 mrem/hr neutrons (Atomics International, 1964a). 

Although neither the instrument nor the methodology is mentioned, an inference may be made about 

the detection level by the two non-zero values. The minimum detection level of the process would be 

a value greater than zero, but at most, 0.5 mrem/hr. 

7.3.1.3 POMR Occupational X-Ray Examinations

NIOSH has found no records indicating that employees at Piqua Organic Moderated Reactor Facility

were required to complete medical examinations, including chest X-rays prior to beginning work, on a 

periodic basis (e.g., annually), or following termination. However, a former Piqua shift supervisor 

indicated that chest X-rays were included in the annual physical (Personal Communication, 2009c). 

Although no records have been identified that indicate that occupational medical X-rays were 

required, the dose associated with X-ray exams can be assessed using the methodology defined in 

ORAUT-OTIB-0006. NIOSH believes that this methodology supports its ability to bound the 

occupational medical X-ray doses for the evaluated class. 

7.3.1.4 Evaluation for Bounding Ambient External Doses during the Operational Period

The ambient environmental external dose is accounted for, and therefore bounded, by the operational 

dose. Film badges were consistently worn. Any environmental exposure would have been 

incidentally monitored along with the occupational exposure. Therefore, further discussion and 

analysis of the external ambient environmental dose will not be included in this report. 

7.3.1.5 Methods for Bounding External Doses during the Operational Period

There is an established protocol for assessing external exposure when performing dose reconstructions 

(these protocol steps are discussed in the following subsections): 

 Photon Dose 

 Beta Dose 

 Neutron Dose 
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 Medical X-ray Dose 

Photon Dose

The photon dose can be assessed based on bounding assumptions from the site summary data for 

external dose (based on information provided in Table 6-1). The photon information, as discussed in 

the personal dosimetry and area monitoring data sections above, supports applying the bounding dose 

of 1 rem for the years 1963 through 1965. 

The upper bound for a particular year would be assigned to individuals who worked at the POMR site. 

The bounding datum for 1966 is for the entire year, although the operational period ends and the postoperational period begins within the year. An employee who worked any portion of 1966 may be 

estimated to have received 2 rem. 

Beta Dose

Survey data reported in the annual and semiannual reports were typically reported as beta-gamma 

(i.e., β-γ). There were a couple of instances where the gamma exposure rate was distinguished from 

the beta-gamma total. These instances described scenarios with the potential for personnel exposure. 

The first instance described a measurement on an in-core filter that had been removed. It had a 

reading of 20 mrad/hr beta-gamma and 5 mR/hr gamma (Atomics International, 1964a, p. 17). The 

second instance was measured as 10 mrads/hr beta-gamma, including 0.5 mR/hr gamma at 1 

inch (Atomics International, 1964c, pp. 19-20). 

The Varskin 3 software application was also used to assess bounding beta-to-gamma ratios for direct 

contact (for hands-on) dose to the extremities, based on the evaluation of source term information. 

Gamma spectroscopy results from three coolant filter media were used as the basis of the analysis; 

cobalt-58, cobalt -60, iron-59, manganese-54, and zinc-65 were identified in the gamma spectroscopy 

results (Atomics International, 1964a, p. 102). The Varskin 3 analysis determined a ratio of 40:1 for 

direct contact with the skin (0 cm air gap) and a 20:1 ratio for a 5 cm air gap (the largest gap allowed 

by Varskin 3). 

For the operational period between 1963 and May 1, 1966, a bounding beta-to-gamma ratio of 40:1 

can be applied when calculating direct contact dose to the hand. This is a bounding value because this 

assumes direct material contact (i.e., coolant) and that no personal protective equipment (e.g., gloves) 

was worn. Furthermore, a bounding beta-to-gamma ratio of 20:1 can be applied to the upper 

extremities (e.g., elbow to hand) or lower extremities (e.g., legs, below the knee), in cases of direct 

contact. The beta-to-gamma ratio of 20:1 for the extremities (excluding the hand) is considered 

bounding because the material would need to be within 5 cm of the body. Furthermore, the extra 

shielding from personal protective equipment is not taken into account; thus, the potential beta 

fractions would be overestimated for the shallow dose calculation. 

Neutron Dose

The potential for neutron dose existed only when the reactor was operating. The potential for neutron 

exposure is not independent of the potential for gamma exposure; that is, there was no place around 

the reactor where personnel could be exposed to neutrons that didn’t also have gamma exposure. 
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Measurements reported in Piqua Nuclear Power Facility Operations Analysis Program Progress 

Report No. 4 (January 1, 1964-June 30, 1964) showed no detectable neutrons in normally-accessible 

areas. In areas where neutrons were detected, the levels were low when compared to the potential for 

gamma exposure. Two values, each indicated as <0.5 mrem/hr, were found. These <0.5 mrem/hr 

readings were taken at the “OC-104, 14” line in the P-1B room” location with a concurrent gamma 

reading of 13.5 mR/hr and at the “OC-101, 14” line in the P-1A room” location with a concurrent 

gamma reading of 11 mR/hr (Atomics International, 1964a, p. 83). 

This survey reported in Piqua Nuclear Power Facility Operations Analysis Program Progress Report 

No. 4 (January 1, 1964-June 30, 1964) was conducted when the plant was operating at full power, and 

should represent the maximum neutron flux. The first detectable reading was associated with a 

gamma level of 13.5 mR/hr, giving a neutron-to-photon ratio of 3.7% (or 0.037 to 1). The second was 

associated with a gamma level of 11 mR/hr, giving a neutron-to-photon ratio of 4.5% (or 0.045 to 1). 

Based on this information, a neutron-to-photon ration of 10% (or 0.10 to 1) can be applied as a 

bounding ratio to assess neutron dose given a photon dose (which would be assigned based the 

bounding photon dose approach described in this section). 

This value is conservative as (1) the actual reported value was “<0.5 mrem/hr”, (2) the survey showed 

no measureable results in the normally occupied areas of the plant, (3) the plant was operating at full 

power, and (4) the ratio is more than twice the value of the measured ratios. In addition, this is 

corroborated by a statement in an interview with a former plant health physics technician who stated 

that he did not recall a positive neutron result on any film badge report and that the only time he 

observed detectable neutron exposure was when calibrating the neutron survey instrument (Personal 

Communication, 2009g). 

Medical X-ray Dose

A dose associated with initial, annual, and termination X-ray exams as part of a routine medical exam 

can be assessed using the methodology defined in ORAUT-OTIB-0006. 

7.3.1.6 External Dose Reconstruction Feasibility Conclusion for the Operational Period

Although no personal dosimetry data were available, NIOSH has access to sufficient information to 

support assessing the external dose for the class evaluated in this report. The method for bounding the 

operational period external dose is based on bounding the doses with the upper limit of the summary 

data for photon dose and through the application of bounding beta-to-gamma and neutron-to-photon 

ratios, which were based on surveys that represent maximum exposure scenarios and source term 

information. Based on this information and the assessment as presented in Section 7.3.1, NIOSH has 

concluded that it is feasible to bound the external dose (reconstruct dose with sufficient accuracy) for 

the operational period for the class evaluated in this report. 

7.3.2 Evaluation of Bounding External Doses during the Post-Operational Period 

The following subsections summarize the extent and limitations of information available for 

reconstructing the external doses during the post-operational period for members of the class under 

evaluation. 
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Due to NIOSH’s finding regarding internal dose (inability to bound), the following subsections are 

included in the event that partial dose reconstructions are deemed necessary. 

7.3.2.1 Personnel Dosimetry Data for the Post-Operational Period

The available external whole-body dosimetry data for the post-operational period are in the form of 

summary data reported to the AEC Chicago Operations for the City of Piqua workers and to the AEC 

Idaho Operations for the Atomics International workers who worked at the POMR facility. The 

summary data cover 1966 through 1968 and are based on film badge data. For external doses during 

the post-operational period, the data appear to be reliable in summary form. The data are presented in 

Table 6-3 of this report. 

In addition to this summary data, there is the Piqua Nuclear Power Facility Final Monthly Progress 

Report No. 69, which describes the termination of the film badge program in 1969 and the condition 

of the plant, which had been remediated to releasable contamination and radiation levels by the 

beginning of that year (Atomics International, 1969). 

Such data would be evaluated for usability on a case-by-case, as needed basis, using NIOSH 

established protocols for dose reconstructions. 

Photon

The upper limit of the ranges reported to the AEC for Piqua workers are 2 rem for the years 1966 and 

1967, and 1 rem for the year 1968 (see Table 6-3 of this report). 

For the year 1969, the data are less direct. However, the plant was in a post-remediation condition 

and the levels were sufficiently low to allow for termination of the film badge program (Atomics 

International, 1969a, p. 11). The maxim direct exposure levels satisfying the contemporary clean-up 

criteria, as verified by survey data in the Piqua Nuclear Facility Final Survey Report, was 0.4 mrad 

per hour beta-gamma (Geiger, 1969). 

Although there were requirements in the planning documents for the use of wrist badges during the 

post-operational period, no results for these badges have been found. 

Beta

Although film badge monitoring in the post-operational period may have included a beta component, 

just as was found for the operational period, no results have been located by NIOSH. However, using 

Varskin 3, a value was developed to provide a beta-to-gamma ratio that reflects the post-operational 

isotopic conditions. 
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7.3.2.2 Area Monitoring Data for the Post-Operational Period

Area monitoring data are available from progress reports, monthly reports, and semiannual reports. 

 

Photon

In the reports from the post-operational period, there are several ambient readings mentioned. Just as 

with the operational period, there are no comprehensive data, only specific measurements listed in the 

text of the reports. Some of these measurements are related to the flushing process and were taken to 

monitor the buildup of the carbonaceous material particulate in filter media. For example, in 

Semiannual Report Number 10, it indicates that radiation levels at the pump strainers "increased from 

8 to 16 mr/hr and from 5 to 10 mr/hr at P-1A and P-1B respectively" (Atomics International, 1967b, p. 

12). 

Other measurements were taken during the post-operational period. For example, the Semiannual 

Report Number 9 indicates a reading of 450 mr/hr at the top of the rotating shield (while the reactor 

was drained), and a reading of 2.2 r/hr without the rotating shield. Although Report Number 9 did 

reference personnel exposure with the statement that " [i]n general, a radiation field of 50 mr/hr 

existed at the periscope operator's location" (Atomics International, 1966e, p. 20), it did not indicate 

what work activities were occurring or what personnel were present, if any. 

The highest readings indicated were taken during the transfer of the upper-grid plate and inner barrel 

from the reactor to the storage pool. Although this is referenced in the Semiannual Report Number 9, 

there is more detail in another report, titled Piqua Nuclear Power Facility Core Unloading and 

Inspection. This report indicates radiation levels during the transfer ranging from "about 500 mr/hr at 

some 20 feet to a high of 20 R/hr at about 2 feet" (Seyfrit, unknown date). The report also included 

some information on personnel dose, indicating that “...exposure was generally limited to the lesser 

rate” and that “the highest individual dose during the transfer was about 120 mr" (Seyfrit, unknown 

date, p. 12). 

NIOSH did not locate any readings in the reports that related to the disassembly and decontamination 

phase. There was, however, a tabulation of beta-gamma contamination readings in the Piqua Nuclear 

Power Facility Final Survey Report that were taken on January 13, 1969 and that indicate the state of 

the plant, post-cleanup (Geiger, 1969). The report listed "fixed (mrad/hr)" and "removable dpm/100 

cm2

" for 79 locations in the plant. The great majority of the fixed readings were 0.02 mrad/hr with the 

highest listed as “0.8 max, 0.1 average" located at the sump wall (Geiger, 1969, p. 7). 

Data from these reports are not in conflict with the data listed in Table 6-3. The data agree, in general 

terms, with the data reported to the AEC and listed in the table as the highest exposures occurred in 

1966 and 1967, and the highest number of employees with doses greater than 1 rem, eight in total, 

occurred in 1966 when the readings were taken during the transfer of the upper-core grid plate and 

inner barrel. 

Beta

There were no readings in the post-operational reports that indicated beta exposure levels or personnel 

doses. Just as with the operational period, the beta dose may be estimated with a beta-to-gamma ratio 

developed from modeling the isotopes expected to be present. 
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7.3.2.3 POMR Occupational X-Ray Examinations

Just as with the operational period, NIOSH has found no requirements for initial, periodic, or 

termination X-rays for the post-operational period. However, following the assumption that medical 

exams, including X-rays, may have been conducted throughout the life of the plant, the dose 

associated with X-ray exams can be assessed using the methodology defined in ORAUT-OTIB-0006. 

NIOSH believes that this methodology supports its ability to bound the occupational medical X-ray 

doses for the evaluated class. 

7.3.2.4 Evaluation for Bounding Ambient External Doses during the Post-Operational Period

The ambient environmental external dose is accounted for by the available post-operational dose data. 

That is, film badges were consistently worn. Any environmental exposure would have been 

incidentally monitored along with the occupational exposure. Therefore, further discussion and 

analysis of the external ambient environmental dose will not be included in this evaluation. 

7.3.2.5 Methods for Bounding External Doses during the Post-Operational Period

There is a protocol for assessing external exposure for the post-operational period when performing 

dose reconstructions (these protocol steps are discussed in the following subsections): 

 Photon Dose 

 Beta Dose 

 Medical X-ray Dose 

Photon Dose

The photon dose can be assessed based on assumptions from the site summary data for external dose 

(based on information provided in Table 6-3). The photon information, as discussed in the personal 

dosimetry and area monitoring data sections above, supports applying 2 rem for the years 1966 and 

1967, and 1 rem for 1968. The upper limit for a particular year would be assigned to individuals who 

worked at the POMR site. 

 

The 2 rem datum for 1966 is for the entire year, although the operational period ends and the postoperational period begins within the year. An employee who worked any portion of the year may be 

estimated to have received 2 rem. 

In 1969, the plant had been remediated to contemporary clean-up criteria by the start of the year. The 

average ambient radiation level for surfaces was to have been at or less than 0.4 mrad/hr when 

measured one centimeter away from any surface (Geiger, 1969, p. 5). 

Beta Dose

Varskin 3 results for the post-operational period show the beta-to-gamma ratio increasing as the 

spectrum becomes increasingly dominated by cobalt-60. For January 1, 1969, the ratio for direct 

contact becomes 59:1 and for the 5 cm air gap, 32:1. 
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Medical X-ray Dose

A dose associated with initial, annual, and termination X-ray exams as part of a routine medical exam 

can be assessed using the methodology defined in ORAUT-OTIB-0006. 

7.3.2.6 External Dose Reconstruction Feasibility Conclusion for the Post-Operational Period

Although NIOSH has not performed an exhaustive evaluation of the potential external exposures from 

post-operations period, NIOSH believes that external dose can be bounded using the methods for the 

operational period. An exhaustive evaluation was not performed since NIOSH has determined that 

internal doses cannot be bounded with sufficient accuracy. NIOSH intends to use the available 

external monitoring data, and dose reconstruction approaches defined for the operational period to 

support partial dose reconstructions (to include medical X-ray exposures) for the post-operational 

period. 

7.4 Evaluation of Petition Basis for SEC-00126 

The petition basis provided in SEC-00126 is that radiation doses potentially incurred by members of 

the proposed class may not have been adequately monitored either through personal monitoring or 

through air monitoring. In support of this basis, the petitioner included the following statement: He 

was a laborer primarily assigned any duty his supervisor would assign. This all done without any 

type of monitoring, training, or protective devices for handling nuclear material this was not 

explained but one has to surmise it was because of [Name Redacted] primary work was at the other 

building. 

NIOSH RESPONSE: Although NIOSH has not located individual data, NIOSH is aware that 

radiological monitoring was performed at the POMR facility. This is evident in the monthly and 

semiannual reports where exposures were reported and summarized. In addition, former POMR 

personnel interviewed for this report stated that they remembered wearing monitoring devices, 

including pencil dosimeters and film badges (Personal Communication, 2009a; Personal 

Communication, 2009c; Personal Communication, 2009d; Personal Communication, 2009e). During 

the operational period, for any potential unmonitored Piqua personnel, it is assumed that exposures 

calculated or measured for monitored Piqua personnel would exceed any potential exposures to 

unmonitored personnel during the operational period. NIOSH is recommending a class be included in 

the SEC for POMR during the covered period from May 2, 1966 through February 28, 1969. 

7.5 Summary of Feasibility Findings for Petition SEC-00126 

This report evaluates the feasibility for completing dose reconstructions for employees at the POMR 

facility from January 1, 1963 through February 28, 1969. NIOSH found that the available monitoring 

records, process descriptions and source term data available are not sufficient to complete dose 

reconstructions for the entire evaluated class of employees. 

Table 7-2 summarizes the results of the feasibility findings at the POMR facility for each exposure 

source during the time period from January 1, 1963 through February 28, 1969. 
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Table 7-2: Summary of Feasibility Findings for SEC-00126 

January 1, 1963 through February 28, 1969 

Operational Period 

(January 1, 1963 through May 1, 1966) 

Post-Operational Period 

(May 2, 1966 through February 28, 1969) 

Source of Exposure Reconstruction 

Feasible 

Reconstruction 

Not Feasible 

Reconstruction 

Feasible 

Reconstruction 

Not Feasible 

Internal1 X X2

Process Related X X 

Ambient/Environmental X X 

External X X3

 Gamma X X 

 Beta X X 

 Neutron X N/A 

 Occupational Medical X-ray X X 

Notes: 

1

 Internal includes an evaluation of reports site, stack, and environmental air concentrations. 

PARTIAL DOSE RECONSTRUCTION INFORMATION: 

2

 INTERNAL: For the post-operational period, any available personnel monitoring data that may exist will be applied on a 

case-by-case basis in support of a partial dose reconstruction for claims with non-presumptive cancers and those with less 

than 250-days of employment. 

3

 EXTERNAL: For the post-operational period, occupational medical X-ray dose, photon dose, and beta dose may be 

estimated in support of a partial dose reconstruction for claims with non-presumptive cancers and those with less than 250-

days of employment. 

As of September 9, 2009, a total of five claims have been submitted to NIOSH for individuals who 

worked at the POMR facility and are covered by the class definition evaluated in this report. Dose 

reconstructions have been completed for three individuals (~60%). 

8.0 Evaluation of Health Endangerment for Petition SEC-00126 

The health endangerment determination for the class of employees covered by this evaluation report is 

governed by both EEOICPA and 42 C.F.R. § 83.13(c)(3). Under these requirements, if it is not 

feasible to estimate with sufficient accuracy radiation doses for members of the class, NIOSH must 

also determine that there is a reasonable likelihood that such radiation doses may have endangered the 

health of members of the class. Section 83.13 requires NIOSH to assume that any duration of 

unprotected exposure may have endangered the health of members of a class when it has been 

established that the class may have been exposed to radiation during a discrete incident likely to have 

involved levels of exposure similarly high to those occurring during nuclear criticality incidents. If 

the occurrence of such an exceptionally high-level exposure has not been established, then NIOSH is 

required to specify that health was endangered for those workers who were employed for a number of 

work days aggregating at least 250 work days within the parameters established for the class or in 

combination with work days within the parameters established for one or more other classes of 

employees in the SEC. 

Based on the absence of monitoring data and the higher exposure potential during the post-operational 

period, NIOSH’s evaluation determined that it is not feasible to estimate radiation dose for all 

members of the NIOSH-evaluated class with sufficient accuracy based on the sum of information 
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available from available resources. Modification of the class definition regarding health 

endangerment and minimum required employment periods, therefore, is required. 

9.0 Class Conclusion for Petition SEC-00126 

Based on its full research of the class under evaluation, NIOSH has defined a single class of 

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class to be added to the SEC includes all employees of the Department of Energy, its 

predecessor agencies, and its contractors and subcontractors who worked at the Piqua Organic 

Moderated Reactor site during the covered period from May 2, 1966 through February 28, 1969, for a 

number of work days aggregating at least 250 work days, occurring either solely under this 

employment or in combination with work days within the parameters established for one or more 

other classes of employees in the Special Exposure Cohort. NIOSH is proposing only the postoperational period based on (1) the more extensive personnel exposure potential during the postoperational period (based on the recovery and decontamination and decommissioning activities 

performed), and (2) the lack of personnel internal monitoring data for the post-operational period. 

NIOSH has carefully reviewed all material sent in by the petitioner, including the specific assertions 

stated in the petition, and has responded herein (see Section 7.4). NIOSH has also reviewed available 

technical resources and many other references, including the Site Research Database (SRDB), for 

information relevant to SEC-00126. In addition, NIOSH reviewed its NOCTS dose reconstruction 

database to identify EEOICPA-related dose reconstructions that might provide information relevant to 

the petition evaluation. 

These actions are based on existing, approved NIOSH processes used in dose reconstruction for 

claims under EEOICPA. NIOSH’s guiding principle in conducting these dose reconstructions is to 

ensure that the assumptions used are fair, consistent, and well-grounded in the best available science. 

Simultaneously, uncertainties in the science and data must be handled to the advantage, rather than to 

the detriment, of the petitioners. When adequate personal dose monitoring information is not 

available, or is very limited, NIOSH may use the highest reasonably possible radiation dose, based on 

reliable science, documented experience, and relevant data to determine the feasibility of 

reconstructing the dose of an SEC petition class. NIOSH contends that it has complied with these 

standards of performance in determining the feasibility or infeasibility of reconstructing dose for the 

class under evaluation. 
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10.0 References 

10 C.F.R. pt. 20, Standards for Protection Against Radiation, Cumulative Supplement of the Federal 

Register; January 1, 1961; SRDB Ref ID: 62588 

42 C.F.R. pt. 81, Guidelines for Determining the Probability of Causation Under the Energy 

Employees Occupational Illness Compensation Program Act of 2000; Final Rule, Federal 

Register/Vol. 67, No. 85/Thursday, p 22,296; May 2, 2002; SRDB Ref ID: 19391 

42 C.F.R. pt. 82, Methods for Radiation Dose Reconstruction Under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule; May 2, 2002; SRDB Ref ID: 

19392 

42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as Members of the Special 

Exposure Cohort Under the Energy Employees Occupational Illness Compensation Program Act of 

2000; Final Rule; May 28, 2004; SRDB Ref ID: 22001 

42 U.S.C. §§ 7384-7385 [EEOICPA], Energy Employees Occupational Illness Compensation 

Program Act of 2000, as amended 

AEC, 1963, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1963, results for Piqua; Atomic Energy Commission (AEC); 1963; 

SRDB Ref ID: 13792, p. 34 

AEC, 1964, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1964, results for Piqua; Atomic Energy Commission (AEC); 1964; 

SRDB Ref ID: 14442, p. 18 

AEC, 1965a, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1965, results for Piqua; Atomic Energy Commission (AEC); 1965; 

SRDB Ref ID: 13797, p. 27 

AEC, 1965b, Radiation Exposure History for One Employee, 1961-1965; U.S. Atomic Energy 

Commission (AEC); 1965; SRDB Ref ID: 62522, p. 3 

AEC, 1966a, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1966, results for Piqua; Atomic Energy Commission (AEC); 1966; 

SRDB Ref ID: 13798, p. 23 

AEC, 1966b, Chicago Operations Office Review of the Phase IIA Recovery Program Plan of Action, 

includes cover letter; Atomic Energy Commission (AEC); June 23, 1966; SRDB Ref ID: 57303, pp. 

275-279 

AEC, 1966c, Authorization to Proceed with Phase I of the AI Piqua Recovery Program, includes 

cover letter; Atomic Energy Commission (AEC); April 15, 1966; SRDB Ref ID: 57624, pp. 31-33 
SEC-00126 09-14-09 FINAL Piqua Organic Moderated Reactor (POMR)

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AEC, 1967, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1967, results for Piqua; Atomic Energy Commission (AEC); 1967; 

SRDB Ref ID: 16915, p. 25 

AEC, 1968, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1968, results for Piqua; Atomic Energy Commission (AEC); 1968; 

SRDB Ref ID: 13799, p. 22 

Arnold, 1966, Irradiation Performance of Piqua Core 1 Fuel Elements, NAA-SR-11712; J. L. 

Arnold; May 25, 1966; SRDB Ref ID: 53719 

Ashley, 1964, Reactor Information, correspondence to J. Jacobson; R. L. Ashley; January 10, 1964; 

SRDB Ref ID: 57974, p. 173 

Ashley, 1966, Phase IIA Procedures, correspondence to W. T. Morgan; R. L. Ashley; June 6, 1966; 

SRDB Ref ID: 57303, p. 422 

Atomics International, unknown date-a, Piqua Nuclear Power Facility Progress Report from 

Criticality to First Scheduled Shutdown (June 1963-May 1964); Atomics International; unknown date; 

SRDB Ref ID: 56012 

Atomics International, unknown date-b, Phase IIA Plan of Action; Atomics International; unknown 

date; SRDB Ref ID: 21870 

Atomics International, 1961, Final Safeguards Summary Report for the Piqua Nuclear Power 

Facility, NAA-SR-5608 Reactor Technology; Atomics International; August 1, 1961; SRDB Ref ID: 

53945, pp. 12-331 

Atomics International, 1963, Piqua Nuclear Power Facility Operations Analysis Program Progress 

Report No. 2 (Fiscal Year 1963), NAA-SR-8722 Reactor Technology; Atomics International; 

November 30, 1963; SRDB Ref ID: 55926 

Atomics International, 1964a, Piqua Nuclear Power Facility Operations Analysis Program Progress 

Report No. 4 (January 1, 1964-June 30, 1964), NAA-SR-10307 Reactor Technology; Atomics 

International; December 15, 1964; SRDB Ref ID: 55814 

Atomics International, 1964b, Piqua Nuclear Power Facility Report of Power Operation; Atomics 

International; March 6, 1964; SRDB Ref ID: 56010 

Atomics International, 1964c, Piqua Nuclear Power Facility Monthly Operating Report No. 19; 

Atomics International; November 1964; SRDB Ref ID: 55945 

Atomics International, 1964d, Piqua Nuclear Power Facility Operations Analysis Program Progress 

Report No. 3 (July 1, 1963-December 31, 1963), NAA-SR-9473 Reactor Technology; Atomics 

International; July 15, 1964; SRDB Ref ID: 55932 
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Atomics International, 1965a, Compilation of Piqua Nuclear Power Facility Operating Limits and 

Controls and Post-Critical Operational Tests; Atomics International; February 5, 1965; SRDB Ref 

ID: 55946 

Atomics International, 1965b, Piqua Nuclear Power Facility Monthly Operating Report No. 24; 

Atomics International; April 1965; SRDB Ref ID: 55943 

Atomics International, 1965c, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 6 (January 1-June 30, 1965); Atomics International; December 15, 

1965; SRDB Ref ID: 55817 

Atomics International, 1965d, Piqua Nuclear Power Facility Monthly Operating Report No. 26; 

Atomics International; June 1965; SRDB Ref ID: 55944 

Atomics International, 1965e, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 5 (July 1-December 31, 1964); Atomics International; June 15, 

1965; SRDB Ref ID: 55815 

Atomics International, 1966a, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 7 (July 1, 1965-January 13, 1966); Atomics International; October 

25, 1966; SRDB Ref ID: 55880 

Atomics International, 1966b, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 8 (January 1-June 30, 1966), NAA-SR-12148 General, 

Miscellaneous; Atomics International; December 31, 1966; SRDB Ref ID: 55881 

Atomics International, 1966c, Piqua Recovery Project Guidance, includes correspondence that 

accompanied the guidance; Atomics International; June 1, 1966; SRDB Ref ID: 57303, pp. 472-474 

Atomics International, 1966d, PNPF Recovery Program Report—Week Ending July 22, 1966; 

Atomics International; July 25, 1966; SRDB Ref ID: 57303, pp. 19-21 

Atomics International, 1966e, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 9 (July 1-December 31, 1966); Atomics International; May 1, 1969; 

SRDB Ref ID: 55883 

Atomics International, 1967a, Piqua Nuclear Power Facility Monthly Operating Report No. 53, COO652-43; Atomics International; September 1967; SRDB Ref ID: 55942 

Atomics International, 1967b, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 10 (January 1-June 30, 1967), NAA-SR-12561 General, 

Miscellaneous; Atomics International; November 1, 1967; SRDB Ref ID: 55892 

Atomics International, 1968a, Draft Piqua Nuclear Power Facility Retirement Plan, AI-AECMEMO-12707; Atomics International; May 30, 1968; SRDB Ref ID: 54873 

Atomics International, 1968b, Piqua Nuclear Power Facility Surveillance and Recovery Program 

Final Report, AI-AEC-12696; Atomics International; June 15, 1968; SRDB Ref ID: 57628 
SEC-00126 09-14-09 FINAL Piqua Organic Moderated Reactor (POMR)

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Atomics International, 1968c, Piqua Nuclear Power Facility Preliminary Study of Safety 

Requirements for Retirement Alternatives; Atomics International; February 16, 1968; SRDB Ref ID: 

57360 

Atomics International, 1968d, Piqua Nuclear Power Facility Retirement Safety Analysis Report, AIAEC-MEMO-12708; Atomics International; August 21, 1968; SRDB Ref ID: 54895 

Atomics International, 1969, Piqua Nuclear Power Facility Final Monthly Progress Report No. 69, 

COO-652-59; Atomics International; February 10, 1969; SRDB Ref ID: 60169 

Cember, 1969, Introduction to Health Physics, 1st edition; Herman Cember; January 1969; SRDB Ref 

ID: Publicly available 

City of Piqua, 1964, 1963 Radiation Exposure Records for AI Personnel at the PNPF, 

correspondence; City of Piqua; February 11, 1964; SRDB Ref ID: 62522, p. 2 

DOE, 2009, Piqua Nuclear Power Facility; DOE Environmental Management website; last accessed 

on March 20, 2009; SRDB Ref ID: 53525 

Fix, 1997, Retrospective Assessment of Personnel Neutron Dosimetry for Workers at the Hanford Site, 

PNNL-11196 UC-606; J. J. Fix, R. H. Wilson, and W. V. Baumgartner; February 1997; SRDB Ref 

ID: 5275 

Form B, 2008, Petition Form B; August 21, 2008; OSA Ref ID: 106811 

Geiger, 1969, Piqua Nuclear Power Facility Final Survey Report; Eric L. Geiger; January 23, 1969; 

SRDB Ref ID: 60170 

Hewson, 1969, PNPF Retirement Safety Analysis Reevaluation of Residual Nuclides, AI-AECMEMO-12708 Supplement A; R. A. Hewson; January 15, 1969; SRDB Ref ID: 58051 

Huntsinger, 1969, Safety Evaluation of PNPF Modifications; M. Huntsinger and R. S. Hart; May 1, 

1969; SRDB Ref ID: 55890 

McCurnin, 1966, Visual Examination of Fuel Elements at PNPF through May 19, 1966, 

correspondence to F. C. Grouemeyer; W. R. McCurnin; SRDB Ref ID: 57303, pp. 25-54 

Morgan, 1966, Piqua Recovery Program—Phase II-A Plan of Action; W. T. Morgan; July 11, 1966; 

SRDB Ref ID: 57303, pp. 104-134 

OCAS-PR-004, Internal Procedures for the Evaluation of Special Exposure Cohort Petitions, Rev. 0; 

National Institute for Occupational Safety and Health (NIOSH); Cincinnati, Ohio; September 23, 

2004; SRDB Ref ID: 32022 

ORAUT-OTIB-0006, Dose Reconstruction from Occupationally Related Diagnostic X-Ray 

Procedures; Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; December 21, 2005; 

SRDB Ref ID: 20220 
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Personal Communication, 2009a, Personal Communication with Former Health Physics Technician; 

Telephone Interview by ORAU Team; January 20, 2009; SRDB Ref ID: 61683 

Personal Communication, 2009b, Personal Communication with DOE Legacy Management 

Employee; Telephone Interview by ORAU Team; January 29, 2009; SRDB Ref ID: 61681

Personal Communication, 2009c, Personal Communication with Former Shift Supervisor; Telephone 

Interview by ORAU Team; February 19, 2009; SRDB Ref ID: 61677

Personal Communication, 2009d, Personal Communication with Former Construction Engineer, 

Instrumentation Engineer, and Electrical Engineer; Telephone Interview by ORAU Team; February 

19, 2009; SRDB Ref ID: 61684 

Personal Communication, 2009e, Personal Communication with Former Reactor Operator and 

Maintenance Foreman; Telephone Interview by ORAU Team; February 6, 2009; SRDB Ref ID: 

61679

Personal Communication, 2009f, Personal Communication with Chief Health Physicist; Telephone 

Interview by ORAU Team; February 23, 2009; SRDB Ref ID: 61680 

Personal Communication, 2009g, Personal Communication with Former Health Physics Technician, 

second interview; Telephone Interview by ORAU Team; March 18, 2009; SRDB Ref ID: 62597

Personal Communication, 2009h, Personal Communication with Former POMR Health Physicist; 

documented telephone communication; March 23, 2009; SRDB Ref ID: 62596 

Personal Communication, 2009i, Personal Communication with Former Shift Supervisor; Telephone 

Interview by ORAU Team; June 23, 2009; SRDB Ref ID: 71375 

Personal Communication, 2009j, Personal Communication with Former Day Shift Supervisor, 

Operations Engineer, and Operations and Maintenance Supervisor; Telephone Interview by ORAU 

Team; June 29, 2009; SRDB Ref ID: 71376 

Personal Communication, 2009k, Personal Communication with Former Health Physicist; Telephone 

Interview by ORAU Team; June 25, 2009; SRDB Ref ID: 71379 

Radiological Health Handbook, 1970, Radiological Health Handbook; U.S. Department of Health, 

Education, and Welfare, compiled and edited by the Bureau of Radiological Health and the Training 

Institute, Environmental Control Administration; January 1970; SRDB Ref ID: publicly available 

Schwartz, 1983, Experimental Verification of the Neutron Spectrum from the NBS D/Sub 2/OModerated/Sup 252/Cf Source, NUREC/CR-3399; R. B. Schwartz, C. M. Eisenhauer, and J. A. 

Grundl; August 1, 1983; SRDB Ref ID: 64008 

Seyfrit, unknown date, Piqua Nuclear Power Facility Core Unloading and Inspection, DOL-08-

04231; K. V. Seyfrit, City of Piqua and J. A. McEdwards, Atomics International; unknown date; 

SRDB Ref ID: 56018 
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Unknown author, unknown date-a, Summary of Operating History, select pages from another 

unidentified document; unknown date; SRDB Ref ID: 21882 

Unknown author, unknown date-b, Piqua Nuclear Power Facility History; unknown author; unknown 

date; SRDB Ref ID: 53751 

Wheelock, 1970, Retirement of the Piqua Nuclear Power Facility, AI-AEC-12832 Reactor 

Technology; C. W. Wheelock; April 1, 1970; SRDB Ref ID: 21860 
SEC-00126 09-11-09 FINAL Piqua Organic 

Attachment One: Information from Interviews with Former Piqua Pe

Table A1-1: Interview Statements Regarding POMR Operational Period 

Dosimetry Survey Neutron General Radiation CAMs Incidents Bioassay 

Operations Shift Supervisor, City of Piqua, July 1960-February/March 1969 (PC, 2009c & PC, 2009i) 

Pencil dosimeters were used during normal operations (Stored in rack when not in use). Film badges were used when fuel came on the site to load the core (until the fuel was put in storage in 1969). 

Routinely No neutrons to be 

concerned about. 

Beta-gamma was the only concern. 

CAMS were not used; the only air monitoring done was in the ventilation system. 

Moderator temperature kept increasing during power operation. [Note: Interviewee thought this was kept hidden, but NIOSH found an entire write up dedicated to this.]

Doesn’t remembeit being performe

HP Technician, City of Piqua, Fall 1961-Spring 1966 (PC, 2009a & PC, 2009g) 

When operations began, all workers wore film badges that were exchanged monthly. Admin workers badges were eventually exchanged quarterly. Of the ~45 workers who were routinely monitored, only 1 or 2 workers would have a positive reading (20-30 mR) above MDL. Control badges were used to determine background levels. Thinks NTA badges used for neutron 

Daily, weekly, and monthly surveys, depending on type of survey. All surveys were documented. Neutron surveys were done routinely, at least monthly. When the power level was raised, extensive neutron surveys were done initially. 

No neutron issues; the greatest neutron dose came from calibrating the neutron survey instrument using a PuBe source. There was a potential for neutron exposure at the primary coolant line (P1B). The levels were less than 0.5 mrem/hr. 

They could do all maintenance work at any time because there was not a dose issue. The only measurable exposure occurred when they were working with the instrument thimbles on top of the reactor in the high-bay area. When tubular fuel rods were pulled, they were about 200 mR/hr. HPs left them on ground, took a coffee break to allow for radioactive decay. 

CAMs operating at all times. Rotating filter media was changed hourly and was read continuously using an end window GM tube. It was all recorded on a strip chart recorder. One sample was taken every month and sent away for outside analysis. A portion of the 

No unusual events that led to either contamination or exposure. 

Does not recall abioassays or WBbeing performed. 

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Table A1-1: Interview Statements Regarding POMR Operational Period 

Dosimetry Survey Neutron General Radiation CAMs Incidents Bioassay 

exposures. Visitors were assigned film badges. As many as 31 workers from Atomics International, Inc. (AI) worked at Piqua because they were responsible for reactor startup. Some of these workers wore both AI and Piqua film badges, but Piqua was responsible for tracking their doses. Dosimetry records first went to the City of Piqua, and then were sent to AEC Chicago Operations Office on a monthly basis. All badges were shipped to Landauer in Chicago. 

Afterwards, the surveys were less extensive. 

Filter change out comment: It was a hot (thermal) and dirty (nonradioactive) job, but the exposure rate was low. The highest gamma dose rate (10-15 mR/h) was around the primary coolant pump coming out of the core. The gamma dose was very small and there was usually nothing to report. No worker had a monthly dose exceeding 50 mR. Cannot remember the highest annual dose, but it was so minimal that it was not a concern. 

filter was sent out monthly for analysis and then sent to AEC Chicago. 

Construction Engineer, Atomics International, 1961-1966 (PC, 2009d) 

All reactor operators and maintenance workers wore film badges all the time, and they also had pencil dosimeters. Recalled that administrative workers had to go to an HP to get a film badge if they were going into the Reactor Building. Years after he left, he 

Contamination surveys were routinely performed; it was a very clean plant. HPs would tell workers where rad levels were. 

Neutrons in core, but no leaks of neutrons from reactor vessel. Containment tight and 6’ of shielding around reactor. 

Was a very, very low-radiation-level place. 

Does not recall any area sampling being performed, but the exhaust air from the stack was monitored. 

Around 1964, in the Reactor Building, a seal on a pump failed and began spewing out organic material coolant. When the seal let loose and the coolant spilled out, supposedly there was a fire, but no one saw it. 

-

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Table A1-1: Interview Statements Regarding POMR Operational Period 

Dosimetry Survey Neutron General Radiation CAMs Incidents Bioassay 

received information from the AEC that told him how much radiation dose he received. He said that the AEC would send exposure records to AI. 

Reactor Operator & Maintenance Foreman, DOE, 1960-1969 (PC, 2009e) 

Does not recall film badge exchange frequency; it may have been done monthly. Never saw the results of film badge measurements, but he received a yearly letter telling him what his exposure was for that year. Does not recall getting any exposure. The pencil dosimeters were used for a whole day and registered a maximum reading of 200 mR. The workers turned the dosimeters over to the HP every day and the HP checked the reading, but he does not know if the readings were recorded. Does not recall ever getting a reading on his dosimeter. Although, he might have gotten a reading on one job when he cleaned up the 

Radiation and contamination surveys were routinely performed. 

No neutron radiation. 

The beta and gamma radiation were from activation of the impurities in the coolant. 

Air monitoring was done inside and outside the POMR plant. 

-

Does not recall giving urine samples. Howevhe was not saying did not happen, jthat he does not remember giving urine samples. 

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Table A1-1: Interview Statements Regarding POMR Operational Period 

Dosimetry Survey Neutron General Radiation CAMs Incidents Bioassay 

residue on the absolute filters from the coolant on a big boiler. 

Health Physicist, City of Piqua (PC, 2009f) 

- Only smear 

surveys of the floors. 

No neutron exposures. 

The highest doses at the site were about 100 mrem in a month. Doses could have also been in the 30-40 mrem/month range. 

No air contamination. 

- - 

Table A1-2: Interview Statements Regarding POMR Post-Operational Period 

Monitoring/Sampling/ 

Surveys 

Shutdown/ Dismantlement/ 

Recovery Activities and 

Shipment 

Radiation 

Materials and/or 

Fission Products 

Coke Removal 

Protective 

Clothing and/or 

Equipment 

Spills/Acciden

Operations Shift Supervisor, City of Piqua, July 1960-February/March 1969 (PC, 2009c & PC, 2009i) 

Radiation surveys were performed routinely. The HPs did smears on surfaces and floors. Workers always had film badges from the time the fuel came on the site to load the core, until the fuel was put into storage in 1969. No CAMs were used. The only air monitoring was done in the ventilation system. Air sampling was performed during dismantlement. 

Piqua employees and some people from Atomics International dismantled the plant. They started cutting the plant up into small pieces with saber saws. Unknown dates— Unloaded everything, drained pool, removed and demolished mass of coke and then wrote the dismantling procedure. 1968—Dismantling began and lasted about a year. Cranes were used to remove parts. However, the fuel handling machine was also used. Dismantled the boiler, super heater, and all other reactor 

No neutrons. Betagamma radiation was the only concern. Does not know if he was ever exposed to any radiation at all. 

All employees were involved. Maintenance crew did the work on the day shift, and the operations staff did the work on the 2nd and 3rd

shifts. Grapplers and manipulators were used to grasp the coke at different angles. Pieces of the coke (including from the upper level, 100) were put into a container that 

Respirators were not required, and do not remember seeing respirators at any time. During the dismantlement process, they wore coveralls unless they were in the administrative area. Only wore street clothes in the reactor area when everything was sealed up. 

Nothing during thpost-operational period. 

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Table A1-2: Interview Statements Regarding POMR Post-Operational Period 

Monitoring/Sampling/ 

Surveys 

Shutdown/ Dismantlement/ 

Recovery Activities and 

Shipment 

Radiation 

Materials and/or 

Fission Products 

Protective 

Coke Removal Clothing and/or Spills/Acciden

Equipment 

components. When the reactor was clean, it was sealed up. All documents were placed on top of the reactor. The spaces were filled in with sand and a concrete floor was put on top of everything. In the lower levels of the reactor building, spilled coolant was removed and areas were cleaned and epoxy paint was used to cover those areas. Size reduction—After the reactor was dismantled, some of the pipes were cut down to size. Some containment and shielding were brought in to ship some of the items. The boiler was lifted onto a truck with a crane and hauled off site. After the dismantlement, materials were stored on the main floor level until it was hauled away on a big truck. 

was located down on the lower level in the reactor, below the work platform. 

Day Shift Supervisor, Operations Engineer, & Operations and Maintenance Supervisor, City of Piqua, July 1965-August 1968 (PC, 2009j) 

Workers wore film badges and there were pocket dosimeters. Has a recollection of urinalysis, but doesn’t recall any whole-body counts being performed. There were CAMs on the 100 foot level and throughout the containment area. 

Piqua employees would have disassembled the reactor. Fuel was removed remotely with a fuel removal apparatus, and nobody was present when it was removed. Sometime December 1967-August 1968—process piping and reactor components were removed and the primary piping was found to be 

There were no radioactive materials (to his knowledge) that workers could have been exposed to. Obviously there were activation and corrosion products. But there were never any 

- Site provided 

coveralls, booties, cotton gloves, and hats. Doesn’t recall seeing respirators. 

Two events, none involving radiatio(1) Seal was lost the primary coolapump. (2) Someone ran truck into the wasupply across the river and shut off the water supply the heat exchangThey used water from a water truc

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Table A1-2: Interview Statements Regarding POMR Post-Operational Period 

Monitoring/Sampling/ 

Surveys 

Shutdown/ Dismantlement/ 

Recovery Activities and 

Shipment 

Radiation 

Materials and/or 

Fission Products 

Protective 

Coke Removal Clothing and/or Spills/Acciden

Equipment 

contamination-free. Piping was cut up by the plant maintenance crew (i.e., welders and mechanics). Believes that the fuel was shipped to the National Reactor Testing Station (now the Idaho National Laboratory) or to the Savannah River Plant. Believes the components also went to the National Reactor Testing Station as well; they were transported by truck. 

indications of a fuel leak. Activation products in the coolant were essentially nonexistent. 

for the heat exchanger. 

Reactor Operator & Maintenance Foreman, DOE, 1960-1969 (PC, 2009e) 

Everyone (except the people working in the office) was monitored by film badges and pencil dosimeters—they were used all of the time. Does not recall providing any urine samples. HP group performed radiation surveys on a regular basis; they took smears on the floors. Air monitoring was performed both inside and outside of the plant. 

All Piqua employees who worked in the plant and knew the equipment worked together on D&D activities. Removed all of the lines (in sections) in the containment buildings and sent them to Kentucky for burial. May/June 1966—Fuel was removed; it wasn’t on site for very long, but does not know where it went. Does not know where the coolant went. A lot of it was burned when they were trying to clean it up. The boiler may have used most of the coolant. 

No neutron radiation. 

- -

Contamination wrarely found; it wfrom coolant spil

Health Physicist, Eberline Instrument Corp. of Santa Fe, 1968-1970 (PC, 2009h) 

Film badges always came back with dose less than the MDL. Landauer Co. analyzed the badges. The highest exposures 

- - - - - 

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Table A1-2: Interview Statements Regarding POMR Post-Operational Period 

Monitoring/Sampling/ 

Surveys 

Shutdown/ Dismantlement/ 

Recovery Activities and 

Shipment 

Radiation 

Materials and/or 

Fission Products 

Protective 

Coke Removal Clothing and/or Spills/Acciden

Equipment 

resulted from calibrating the instruments. Urinalysis was done on employees once per year. Eberline analyzed the samples. 

Health Physicist, City of Piqua, early 1966-June 1968 (PC, 2009k) 

Eberline provided and processed the workers’ dosimeters. Was monitored and do not recall any film badge readings over 100 mR. Most readings were zero. Does not recall any worker getting over 1 rem in any year. Radiation levels around the primary coolant piping were less than 100 mrem/h when the reactor was operating. There could have been levels over 100 mrem/h when the piping was cut out, there could have been some internal exposure. No recollection of any internal monitoring being performed. Job specific monitoring/air sampling was not performed. Does not remember any air sampling being performed, but does remember contamination surveys being done. 

Most everyone working the day shift participated in the dismantlement, with an HP generally being present. Does not think 2nd or 3rd shift staff 

did any decommissioning. Operators did not work directly over the reactor. The fuel handling machine was used for some of these activities, including removing the fuel. Fuel that was removed was moved to the fuel storage area and shipped to the Savannah River site. Workers were around part of the time when they were handling fuel. Between December 1967 and February 1969—They removed most of the primary coolant piping and cleaned up and removed other piping. Some was put in the reactor vessel and some was shipped off. The only material he remembers shipping offsite was low level waste in 55-gallon drums to Maxey Flats, Kentucky. 

-

As the HP, involved in monitoring the activities during this phase, but was present only part of the time that the activities were going on. Coke that was removed was probably left in the reactor vessel. 

Wore regular, white coveralls (from Defense Apparel) when working in a containment area in the reactor area. Does not remember requiring respirators for any job, but it is likely that workers wore half-face respirators when cutting out piping. 

There were no accidents or significant spills involving radioactive materials. Occasionally, thewas a leak from a valve. 

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Attachment Two: Data Capture Synopsis 

Table A2-A: Summary of Holdings in the SRDB for POMR 

Data Capture Information Data Captured Description Comple

Primary Site/Company Name: Piqua Organic Moderated Reactor; DOE 1963-1966 Other company names:

Piqua Nuclear Power Facility PNPF Atomics International, now Boeing at Santa Field Laboratory, 1963-1964, see below for data capture results 

See below. See below 

State Contacted: Ohio Department of Health, David Lipp 

No relevant documents not already in the SRDB identified. 11/03/2008 

American Municipal Power-Ohio, Kent Carson, Sr. No relevant documents identified. 02/02/2009 Cincinnati Public Library An operational report from the first criticality to the first scheduled 

shutdown. 

11/06/2008 

Claimant Documented communications, monthly report #69, and a final survey 

report. 

02/25/2009 

Comprehensive Epidemiologic Data Resource (CEDR) 

No relevant documents identified. 10/10/2008 

Department of Labor (Paragon) Testing and operation reports at 50% and 100% power, core unloading and 

inspection report and drawings, facility photos, failure analyses, and 

cathodic protection of the containment shell. 

12/05/2008 

DOE Hanford Declassified Document Retrieval System (DDRS) 

No relevant documents identified. 10/10/2008 

DOE Legacy Management Considered Sites All Considered Sites documents are contained in the Grand Junction Office 

DVD's. No search was necessary. 

N/A 

DOE Legacy Management - Grand Junction Office Annual site radiological surveys, FUSRAP documents, total residual 

radioactivity in the reactor, building drawings, and remediation planning 

documents. 

11/24/2008 

DOE OpenNet No relevant documents identified. 10/10/2008 DOE OSTI Energy Citations Reports on shipping casks and waste treatment, and an Energy Research and 

Development Administration Decontamination and Decommissioning conference. 

01/14/2009 

DOE OSTI Information Bridge A 1991 annual site inspection and radiological survey, reports on the 

performance and metallurgy of organic moderated reactor fuel elements, 

and a study of military applications. 

01/14/2009 

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Table A2-A: Summary of Holdings in the SRDB for POMR 

Data Capture Information Data Captured Description 

Comple

Google FUSRAP site fact sheet, a discussion of entombment as a decommissioning 

technology, AEC and DOE reports, and a 2008 annual site inspection and radiological survey. 

10/25/2008 

Interlibrary Loan Facility engineering design and a status report. 12/19/2008 National Academies Press (NAP) No relevant documents identified. 10/10/2008 National Nuclear Security Administration (NNSA) - Nevada Site Office 

No relevant documents identified. 10/10/2008 

NRC Agencywide Document Access and Management (ADAMS) 

No relevant documents identified. 10/10/2008 

Office of Scientific & Technical Information (OSTI) 

Operations analysis progress reports, engineering studies of organic moderation, and information on preparation and shipping of fuel elements. 

12/16/2008 

Piqua Public Library No relevant documents identified. 01/27/2009 Science Applications International Corporation (SAIC) 

Complex-wide exposure records for 1960 and 1961. 09/02/2004 

San Bruno Federal Records Center Quality assurance documentation and weekly reports, all relating to the deactivation of the facility. 02/02/2006 Santa Susanna Field Laboratory (Boeing) Fuel element safety analysis, fuel shipment reports, criticality controls, 

radiological surveillance program, recovery plans, weekly highlights reports, studies of organic contaminant formation, 1967 recovery program plans, coolant surveillance reports, reactor operations analysis program reports, control rod studies, modification studies, coolant chemistry reports, coolant filtration studies, core examination reports, and facility retirement 

reports. 

12/05/2008 

Unknown Personnel exposure information for 1963 and film badge records for 2 

employees. 

07/31/2003 

Washington State University (U.S. Transuranium and Uranium Registries) 

No relevant documents identified. 10/10/2008 

TOTAL 

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80 of 89 

Table A2-B: Database Searches for POMR 

Database/Source Keywords Hits 

DOE OpenNet http://www.osti.gov/opennet/advancedsearch.jsp COMPLETED 10/10/2008 

"Piqua" + 01/01/1960 - 08/10/2008 9 

DOE Hanford DDRS http://www2.hanford.gov/declass/ COMPLETED 10/10/2008 

"Piqua" + 01/01/1960 - 08/10/2008 0 

DOE OSTI Energy Citations http://www.osti.gov/energycitations/ COMPLETED 01/14/2009 

"Piqua" + Reactor + 01/01/1960 - 08/10/2008 194 

DOE OSTI Information Bridge http://www.osti.gov/bridge/advancedsearch.jsp COMPLETED 01/14/2009 

"Piqua" + 01/01/1960 - 08/05/2008 118 

DOE CEDR http://cedr.lbl.gov/ COMPLETED 10/10/2008 

"Piqua" 0 

NNSA - Nevada Site Office www.nv.doe.gov/main/search.htm COMPLETED 10/10/2008 

"Piqua" 0 

U.S. Transuranium & Uranium Registries http://www.ustur.wsu.edu/ COMPLETED 10/10/2008 

"Piqua" 0 

National Academies Press http://www.nap.edu/ COMPLETED 10/10/2008 

"Piqua" 0 

NRC ADAMS Reading Room http://www.nrc.gov/reading-rm/adams/webbased.html COMPLETED 10/10/2008 

"Piqua" + 01/01/1960 - 08/10/2008 73 

Google http://www.google.com COMPLETED 10/25/2008 

americium OR Am241 OR Am-241 OR "AM 241" OR 241Am OR 241-Am OR "241 Am" AND "Piqua" ionium OR Th230 OR Th-230 OR "Th 230" OR 230Th OR 230-Th OR "230 Th" AND "Piqua" neptunium OR Np237 OR Np-237 OR "Np 237" OR 237Np OR 237-Np OR "237 Np" AND "Piqua" 

7,435 
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Table A2-B: Database Searches for POMR 

Database/Source Keywords 

Hits 

polonium OR Po210 OR Po-210 OR "Po 210" OR 210Po OR 210-Po OR "210 Po" AND "Piqua" thorium OR Th232 OR Th-232 OR "Th 232" OR 232Th OR 232-Th AND "Piqua" AND Reactor "232 Th" OR "Z metal" OR myrnalloy OR "chemical 10-66" OR "chemical 10-12" AND "Piqua" ionium OR UX1 OR UX2 OR Th-230 OR Th230 OR "Th 230" OR 230-Th OR "230 Th" AND "Piqua" 230Th OR Th-234 OR Th234 OR "Th 234" OR 234-Th OR 234Th OR "234 Th" AND "Piqua" tritium OR H3 OR H-3 OR mint OR HTO AND "Piqua" AND Reactor uranium OR U233 OR U-233 OR "U 233" OR 233U OR 233-U OR "233 U" AND "Piqua" AND Reactor U234 OR "U 234" OR U-234 OR 234U OR 234-U OR "234 U" AND "Piqua" U235 OR "U 235" OR U-235 OR 235-U OR 235U OR "235 U" OR U238 AND "Piqua" "U 238" OR U-238 OR 238-U OR 238U OR "238 U" AND "Piqua" U308 OR "U 308" OR U-308 OR 308-U OR 308U OR "308 U" OR "uranium extraction" OR "black oxide" OR "brown oxide" AND "Piqua" "green salt" OR "orange oxide" OR "yellow cake" OR UO2 OR UO3 AND "Piqua" UF4 OR UF6 OR C-216 OR C-616 OR C-65 OR C-211 OR U3O8 AND "Piqua" AND Reactor plutonium OR Pu-238 OR Pu238 OR "Pu 238" OR 238Pu OR 238-Pu OR 

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Table A2-B: Database Searches for POMR 

Database/Source Keywords 

Hits 

"238 Pu" AND "Piqua" AND Reactor Pu-239 OR Pu239 OR "Pu 239" OR 239Pu OR 239-Pu OR "239 Pu" AND "Piqua" Pu-240 OR Pu240 OR "Pu 240" OR 240Pu OR 240-Pu OR "240 Pu" AND "Piqua" Pu-241 OR Pu241 OR "Pu 241" OR 241Pu OR 241-Pu OR "241 Pu" AND "Piqua" radium OR Ra-226 OR Ra226 OR "Ra 226" OR 226-Ra OR 226Ra OR 226-Ra AND "Piqua" AND Reactor Ra-228 OR Ra228 OR "Ra 228" OR 228Ra OR 228-Ra OR "228 Ra" AND "Piqua" radon OR Rn-222 OR Rn222 OR "Rn 222" OR 222Rn OR 222-Rn OR "222 Rn" AND "Piqua" AND Reactor thoron OR Rn-220 OR Rn220 OR "Rn 220" OR 220Rn OR 220-Rn OR "220 Rn" AND "Piqua" AND Reactor protactinium OR Pa-234m OR Pa234m OR "Pa 234m" OR 234mPa OR 234m-Pa OR "234m Pa" AND "Piqua" AND Reactor strontium OR Sr-90 OR Sr90 OR "Sr 90" OR 90-Sr OR 90Sr OR "90 Sr" AND "Piqua" AND Reactor oralloy OR postum OR tuballoy OR "uranyl nitrate hexahydrate" OR UNH OR K-65 OR "sump cake" AND "Piqua" AND Reactor "uranium dioxide" OR "uranium tetrafluoride" OR "uranium trioxide" AND "Piqua" "uranium hexafluoride" OR "air count" AND "Piqua" accident AND "Piqua" AND Reactor 

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Table A2-B: Database Searches for POMR 

Database/Source Keywords 

Hits 

"air dust" OR "air filter" OR "airborne test" AND "Piqua" AND Reactor "alpha particle" OR "belgian congo ore" OR bioassay OR bio-assay AND "Piqua" AND Reactor breath OR "breathing zone" OR BZ OR calibration OR "chest count" OR collimation OR columnation AND "Piqua" AND Reactor contamination OR curie OR denitration OR "denitration pot" AND "Piqua" AND Reactor derby OR regulus OR dose OR dosimeter AND "Piqua" AND Reactor dosimetric OR dosimetry OR electron OR environment AND "Piqua" AND Reactor "Ether-Water Project" OR exposure OR "exposure investigation" OR "radiation exposure" AND "Piqua" AND Reactor external OR "F machine" OR fecal OR "feed material" OR femptocurie OR film OR fission OR fluoroscopy AND "Piqua" AND Reactor "Formerly Utilized Sites Remedial Action Program" OR FUSRAP OR gamma-ray OR "gas proportional" OR "gaseous diffusion" AND "Piqua" AND Reactor health OR "health instrument" OR "health physics" OR "H.I." OR HI OR HP OR "highly enriched uranium" OR HEU AND "Piqua" AND Reactor hydrofluorination OR "in vitro" OR "in vivo" OR incident OR ingestion OR inhalation OR internal AND "Piqua" AND Reactor investigation OR isotope OR isotopic OR "isotopic enrichment" OR "JS Project" OR Landauer OR "liquid scintillation" AND "Piqua" AND Reactor log OR "log sheet" OR "log book" OR "low enriched uranium" OR LEU AND "Piqua" AND Reactor 

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Table A2-B: Database Searches for POMR 

Database/Source Keywords 

Hits 

"maximum permissible concentration" OR MPC OR metallurgy OR microcurie OR millicurie AND "Piqua" AND Reactor "mixed fission product" OR MFP OR monitor OR "air monitoring" OR nanocurie OR "nasal wipe" OR neutron OR "nose wipe" AND "Piqua" AND Reactor nuclear OR Chicago-Nuclear OR "nuclear fuels" OR "nuclear track emulsion" OR "type A" AND "Piqua" AND Reactor NTA OR "occupational radiation exposure" OR occurrence OR "ore concentrate" OR "PC Project" AND "Piqua" AND Reactor permit OR "radiation work permit" OR "safe work permit" OR "special work permit" OR RWP OR SWP AND "Piqua" AND Reactor "phosphate research" OR photofluorography OR photon OR picocurie OR pitchblende OR "pocket ion chamber" OR PIC OR problem OR procedure AND "Piqua" AND Reactor radeco OR radiation OR radioactive OR radioactivity OR radiograph OR radiological AND "Piqua" AND Reactor "Radiological Survey Data Sheet" OR RSDS OR radionuclide OR raffinate OR reactor AND "Piqua" AND Reactor respiratory OR "retention schedules" OR roentgen AND "Piqua" AND Reactor sample OR "air sample" OR "dust sample" OR "general area air sample" AND "Piqua" AND Reactor "solvent extraction" OR source OR "sealed source" OR spectra OR spectrograph OR spectroscopy AND "Piqua" AND Reactor spectrum OR standard OR "operating standard" OR "processing standard" AND "Piqua" AND Reactor 

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Table A2-B: Database Searches for POMR 

Database/Source Keywords 

Hits 

survey OR "building survey" OR "routine survey" OR "special survey" OR "technical basis" AND "Piqua" AND Reactor "thermal diffusion" OR "thermoluminescent dosimeter" OR TLD OR "Tiger Team" AND "Piqua" AND Reactor "tolerance dose" OR urinalysis OR urine OR "whole body count" OR WBC AND "Piqua" AND Reactor "working level" OR WL OR X-ray OR "X ray" OR Xray OR "x-ray screening" AND "Piqua" AND Reactor "Atomics International" AND "Piqua" DOE/CH-9111 OR DOE/CH-9225 

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Table A2-C: OSTI Documents Ordered Regarding POMR 

Document Number Document Title Request

N/A OSTI ID: 5179961 SRDB: 54696 

Piqua Nuclear Power Facility Radiological Surveillance Program 10/29/20

N/A OSTI ID: 4666935 SRDB: 55945 

Piqua Nuclear Power Facility Monthly Operating Report No. 19 10/29/20

N/A OSTI ID: 4614403 SRDB: 55944 

Piqua Nuclear Power Facility Monthly Operating Report No. 26 10/29/20

N/A OSTI ID: 4592064 SRDB: 55943 

Piqua Nuclear Power Facility Monthly Operating Report No. 24 10/29/20

N/A OSTI ID: 4181116 SRDB: 55942 

Piqua Nuclear Power Facility Monthly Operating Report No. 53 10/29/20

NAA-SR-12561 OSTI ID: 4501434 SRDB: 55892 

Piqua Nuclear Power Facility Reactor Operations Analysis Program Semiannual Progress 

Report No. 10, January 1--June 30, 1967. 

10/29/20

NAA-SR-12445 OSTI ID: 4799966 SRDB: 55883 

Piqua Nuclear Power Facility Reactor Operations Analysis Program Semiannual 

Progress Report No. 9, July 1--December 31, 1966. 

10/29/20

NAA-SR-MEMO-12540 OSTI ID: 4496873 SRDB: 55938 

Piqua Nuclear Power Facility Reactor Operations Analysis Program Report 55, June 1967. 

10/29/20

NAA-SR-4576 OSTI ID: 4143395 SRDB: 55904 

Piqua OMR Waste Gas Disposal 10/29/20

NAA-SR-8722 OSTI ID: 4123254 SRDB: 55926 

Piqua Nuclear Power Facility Operations Analysis Program Progress Report No. 2, Fiscal Year 1963 

10/29/20

NAA-SR-5688 OSTI ID: 4080480 SRDB: 55906 

Proceedings of the Organic Cooled Reactor Forum, October 6-7, 1960 10/29/20

NAA-SR-11628 OSTI ID: 4565167 SRDB: 55817 

Piqua Nuclear Power Facility Reactor Operations Analysis Program Semiannual Progress 

Report No. 6, January 1-June 30, 1965 

10/29/20

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Table A2-C: OSTI Documents Ordered Regarding POMR 

Document Number Document Title Request

NAA-SR-9473 OSTI ID: 4031136 SRDB: 55932 

Piqua Nuclear Power Facility Operations Analysis Program Progress Report No. 3, July 1, 1963-December 31, 1963 

10/29/20

NAA-SR-MEMO-10529 OSTI ID: 4639851 SRDB: 55934 

Preparation and Transportation of Piqua Fuel Elements. Summary Report 10/29/20

NAA-SR-11142 OSTI ID: 4583353 SRDB: 55815 

Piqua Nuclear Power Facility Reactor Operations Analysis Program Semiannual Progress 

Report No. 5, July 1-December 31, 1964 

10/29/20

NAA-SR-11650 Sect 3 OSTI ID: 4582815 SRDB: 55818 

Hallam Nuclear Power Facility and Piqua Nuclear Power Facility 10/29/20

NAA-SR-10307 OSTI ID: 4683534 SRDB: 55814 

Piqua Nuclear Power Facility Operations Analysis Program Progress Report No. 4, January 1-June 30, 1964 

10/29/20

NAA-SR-MEMO-12071 OSTI ID: 4468274 SRDB: 55935 

Piqua Nuclear Power Facility Reactor Operations Analysis Program Report No. 41, 

January--April 1966 

10/29/20

NAA-SR-11995 OSTI ID: 4465693 SRDB: 55880 

Piqua Nuclear Power Facility Reactor Operations Analysis Program Semiannual Progress 

Report No. 7, July 1, 1965--January 13, 1966 

10/29/20

NAA-SR-12148 OSTI ID: 4440848 SRDB: 55881 

Piqua Nuclear Power Facility Reactor Operations Analysis Program Semiannual Progress 

Report No. 8, January 1--June 30, 1966 

10/29/20

NAA-SR-11200 OSTI ID: 4571317 SRDB: 55816 

Aec Unclassified Programs. Quarterly Technical Progress Report, January-March 1965 10/29/20

NAA-SR-3575 NAA-SR-MEMO-4048 OSTI ID: 4182273 SRDB: 55896 & 55940 

Preliminary Safeguards Report for the Piqua Organic Moderated Reactor (revised) 

Supplement I to the Preliminary Safeguards Report for the Piqua Organic Moderated Reactor (revised) 

10/29/20

NAA-SR-6893 OSTI ID: 4781060 SRDB: 55921 

Energy Absorption Rates in OMR Coolants 10/29/20

NAA-SR-MEMO-6581 OSTI ID: 4803723 SRDB: 55941 

Re-evaluation of Energy Absorption Rate in OMR`s 10/29/20

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Table A2-C: OSTI Documents Ordered Regarding POMR 

Document Number Document Title Request

NAA-SR-12516 OSTI ID: 4799937 SRDB: 55890 

Safety Evaluation of PNPF Modifications 10/29/20

NAA-SR-3635 OSTI ID: 4221122 SRDB: 55901 

Preliminary Design Description for the Piqua Organic Moderated Reactor Plant, Piqua, Ohio (revised) 

10/29/20

N/A OSTI ID: 4343710 SRDB: 56036 

Recovery of Uranium from Piqua Cold Scrap Fuel Elements 10/29/20

DPSPU-72-124-5 OSTI ID: 4602683 SRDB: 55811 

Safety Analysis Report: Piqua--Elk River Shipping Cask (packaging of radioactive and 

fissile materials). Final Report. 

10/29/20

N/A OSTI ID: 4498453 

Piqua Nuclear Power Facility Surveillance and Recovery Program. Final Report 10/29/20

N/A OSTI ID: 4683116 

Piqua Elk River Reactors Spent Fuel Shipping Cask Design Report 10/29/20

N/A OSTI ID: 5446994 

Annual Survey of the Retired Piqua Nuclear Power Facility 10/29/20

DOE/CH/10370-1 OSTI ID: 6355044 

Annual Survey of the Retired Piqua Nuclear Power Facility 10/29/20

TID-18813 OSTI ID: 4702649 

Piqua-Elk River Reactors Spent Fuel Shipping Cask. Final Design Report 10/29/20

NAA-SR-Memo-7676 OSTI ID: 4728186 

Piqua Operations Analysis Program. Annual Technical Progress Report, Fiscal Year 1962 10/29/20

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Table A2-D: Cincinnati Public Library Documents Ordered for POMR 

Document Number Document Title Request

N/A OSTI ID: 4815595 SRDB: 56595 

Piqua Nuclear Power Facility. 1. Status Report 2. General Description of the Piqua Reactor 

10/29/2008 

N/A OSTI ID: 4196389 SRDB: 56588 

Engineering Design of Piqua OMR 10/29/2008 

N/A OSTI ID: 4676966 SRDB: 53465 

Operating of Piqua Nuclear Power Facility from Criticality to First Scheduled Shutdown 10/29/2008 

N/A OSTI ID: 4147910 

Safety Considerations for the Organic Moderated Reactor in Piqua, Ohio 10/29/2008